International Conference on Advancing the Global Implementation of Decommissioning and Environmental Remediation Programmes CN-238

23 May 2016, 08:00 → 27 May 2016, 18:00 Europe/Madrid

Madrid

Much remains to be done in terms of addressing the legacies from the early development of nuclear energy, including the dismantling of redundant research and fuel cycle facilities, research reactors and nuclear power plants, and the remediation of former nuclear sites and those affected by past uranium mining and processing operations, by other activities involving the use of naturally occurring radioactive material (NORM) or by major nuclear or radiological accidents. Long term solutions often still need to be found for management of the resulting waste, including the development of disposal facilities that meet public acceptance and safety requirements. Some countries are moving forward with dealing with these legacies, and accordingly have built up appropriate technical resources and expertise, but many national programmes still face very significant challenges.

The last major conferences organized by the International Atomic Energy Agency (IAEA) relevant to the above topics took place in Athens, Greece (2006: decommissioning) and in Astana, Kazakhstan (2009: environmental remediation). In the meantime, significant developments in these areas have taken place in some countries and therefore it is timely to discuss the implications for relevant programmes and to exchange recent experiences on these topics. The proposal to combine the two subjects in one conference is a recognition that significant synergies exist between the two activities, which should be explored to foster and optimize the implementation of both decommissioning and environmental remediation worldwide.

The conference will share and review challenges, achievements and lessons learned related to the decommissioning and environmental remediation programmes that have been implemented during the past decade. Key goals will include raising awareness of the importance of addressing the legacies from past activities, identifying current priority needs and providing recommendations on the strategies and approaches that can enable and enhance safe, secure and cost-effective implementation of national and international programmes during the next one to two decades.

Form A http://www-pub.iaea.org/MTCD/Meetings/PDFplus/2016/cn238/cn238FormA.pdf

Form B http://www-pub.iaea.org/MTCD/Meetings/PDFplus/2016/cn238/cn238FormB.pdf

Form C http://www-pub.iaea.org/MTCD/Meetings/PDFplus/2016/cn238/cn238FormC.pdf

Synopsis sample DER2016_template.docx

Monday, 23 May

08:00 → 09:00 Registration of Delegates

09:00 → 11:00 Conference Opening Session 1

09:00 Welcome remarks on behalf of the host country

11:00 → 11:30 Coffee break

11:30 → 13:30 Conference Opening Session 2

12:00 Decommissioning and Remediation in the Russian Federation: Main Results and Future Plans

For a long time nuclear power and nuclear industry in Russia were operated under planned economy and state ownership. It was assumed that RW management and decommissioning activities were to be performed according to a predefined plan using state funds. Changes in the economic management system caused the accumulation of significant challenges in decommissioning and SNF and RW management. By mid-2000, the challenges facing nuclear industry were mainly associated with:  over 150 shut-down nuclear and radiation hazardous facilities awaiting decommissioning, including 4 NPP units and 10 uranium production reactors;  engineered barrier systems at certain nuclear and radiation hazardous facilities some of which operated for more than 50-60 years required urgent overhaul, including water reservoirs (Karachay, the Techa Cascade of water reservoirs, settling ponds) and tailings.  accumulated SNF inventory that totaled 18 500 tons while SNF storage capacity at RBMK and EPG-6 reactor sites was almost exhausted;  need for new disposal facilities designed for different RW classes;  lack of a clear straightforward legal framework to address nuclear legacy challenges;  absence of RW management and decommissioning funds;  absence of real market providing engineering services in RW management and decommissioning. In 2008, the federal target program “Nuclear and Radiation Safety in 2008-2015” was launched. It was carried out in parallel with conversion of nuclear entities into joint stock companies and the establishment of new requirements in RW and SNF management and decommissioning including certain legal standards, thus seeking to avoid further accumulation of issues. In 2015, the federal target program was completed and the Government of the Russian Federation approved a similar program for 2016-2030. The State Corporation “Rosatom” was appointed as the main contractor. The report discusses the key milestones in the program deployment as well as the main results of nuclear legacy decommissioning and cleanup efforts. The primary focus is on: - issues associated with safe retrieval of SNF and nuclear materials from nuclear facilities being part of pre-decommissioning efforts. Over 25,000 SFAs (RBMK-1000, AMB, all types of SNF from research reactors and nuclear-powered ships) were either delivered to centralized storage facilities or reprocessed under the program; - legal framework development, establishment of a state RW management system covering the procedures for inventorying the accumulated RW, creation of new RW disposal facilities. The state of over 2,000 nuclear facilities has been assessed countrywide. In 2016-2030, relevant efforts are scheduled for many of them (the most hazardous ones). Based on integrated risk-cost analysis, strategic decisions have been justified for all legacy RW, practical solutions have been identified for many RW sites. - actual situation in decommissioning. Under the program, efforts have been performed at 200 facilities, 53 of which have already been decommissioned, including a production uranium-graphite reactor, a number of nuclear fuel cycle facilities and research complexes; - remediation of contaminated territories. 4,259,000 m2 of contaminated lands have been remediated under the program; - establishment of centers for decommissioning excellence; - public attitude towards the nuclear back-end activities; - current progress in nuclear submarines dismantlement and remediation of coastal bases. These efforts have been carried out for 15 years already, also under the Global Partnership Initiative; - work plans for 2016 – 2030.

Speaker: Mr Oleg Kryukov (State Corporation “Rosatom”)

13:30 → 15:00 Lunch break

15:00 → 18:15 Session 1 - Poster

15:00 NATIONAL POLICY AND NATIONAL PROGRAM OF SLOVAKIA IN DECOMMISSIONING AND SPENT FUEL MANAGEMENT

The content of the National policy and National program for handling of spent nuclear fuel and radioactive waste in Slovakia was stated according to provision of the EC Directive 70/2011 and Slovak Act No. 238/2006. Both documents were written by Board of Governors of National Nuclear Fund of SR. The National Policy is found on following policies: a) The Slovak Republic shall bear final responsibility for decommissioning of the nuclear installations at the territory of SR, for safe and responsible long-term storage and disposal of spent nuclear fuel and for handling of radioactive waste, which will be produced at its territory after its release by producer by expiring 12 month period from its generation, b) final responsibility for safe and responsible disposal of radioactive waste or spent nuclear fuel, which will be transported from the The Slovak Republic for conditioning or processing to member state of The European Union or the third state included any waste, which will be produced as a by-product of conditioning or processing shall be borne by The Slovak Republic bound by international treaty subject to the provisions, c) generation of radioactive waste from the viewpoint of its activity and volume is maintained at the lowest possible achievable level through appropriate project measures and operational procedures and decommissioning procedures included processing and reuse of materials, d) in all steps of spent nuclear fuel and radioactive waste handling mutual inter dependencies are considered, e) spent nuclear fuel and radioactive waste handling must be safe from long-term viewpoint also, when especially passive safety features are applied, f) in spent nuclear fuel and radioactive waste handling graded approach is applied especially considering activity, amount, type of nuclear facility in which handling is executed and other dangerous characteristics, g) costs for spent nuclear fuel and radioactive waste handling shall be borne by producer, who produced them, in case of unknown originator relevant measures are adopted, h) documentation of decision making process is based on evidence and characterization results in all stages for handling of spent nuclear fuel and radioactive waste. The Government of Slovak Republic approved the National policy and national program of Slovakia in decommissioning and spent fuel management at Jul 27, 2015 via decision No 387/2015. Safe and reliable decommissioning of nuclear installations in the Slovak Republic is achieved throughout application of legislative requirements and also corresponding license and inspection processes provided for in the Atomic Act and its implementing regulations.

Speaker: Prof. Vladimir Slugen (FEI STU Bratislava, Ilkovicova 3, Sk-81219, Slovakia)

15:00 POLICY AND STRATEGIES FOR DECOMMISSIONING AND ENVIRONMENTAL REMEDIATION

Environmental remediation is fundamentally different from radioactive waste management and decommissioning of nuclear installations in that the radioactive material of concern is mixed and/or incorporated into the natural environmental media. However, the remediation policy and strategies need to be coherent and consistent with those of decommissioning and waste management. In general, environmental remediation is applied to the near surface terrestrial environment (rather than, for example, to airborne radioactive substances). Developing a national policy and an underlying strategy or strategies to implement environmental remediation is therefore imperative if the problem holders and decision makers are to succeed in applying the most appropriate and sustainable solutions to their environmental problems. It is important that the Government policy on the land contamination should be built around the twin ideas of stopping contamination of land while taking a risk based approach to tackling historical contamination.

Speaker: Mr Khaled DEBBABI (Head of Service in the Tunisian Nuclear Power Plant Project)

15:00 Strategic Considerations for the Sustainable Remediation of Nuclear Installations

Within the Nuclear Energy Agency (NEA) of the Organisation for Economic Co-operation and Development (OECD), the Working Party on Decommissioning and Dismantling (WPDD) provides a focus for the analysis of decommissioning policy, strategy and regulation. Beyond policy and strategy considerations, the WPDD also reviews practical considerations for implementation such as techniques for characterisation of materials, for decontamination and for dismantling. The Task Group on Nuclear Site Restoration of the WPDD has been given the role of providing the member countries with up-to-date information and to develop consensus regarding strategic aspects of site remediation at nuclear installations. It is discipline oriented, comprising experts in the field of nuclear site remediation, policy makers, regulators, and implementers nominated by the member organisations of the WPDD. To help achieve this task it will keep under review relevant worldwide experience approaches and constraints on the remediation of nuclear sites, management of land and groundwater that are affected by radiological, chemical and/or hazardous materials contamination. The report to be produced in the Spring of 2016 will contain observations and recommendations to consider in the development of strategies and plans for site remediation at nuclear sites to support on-going and new projects to achieve improvements in land quality management consistent with best practice.

Speaker: Mr Peter Orr (Environment Agency)

15:00 THE CHORNOBYL NPP EXCLUSION ZONE 30 YEARS AFTER THE ACCIDENT: LOOKING INTO FUTURE

After the permanent shutdown of the last operating unit of the Chornobyl NPP in December 2000 extensive activities had been started to develop a proper radwaste and spent fuel management infrastructure for the Chornobyl NPP decommissioning and to ensure a long-term solution for spent fuel and for the destroyed Chornobyl-4. Eventually a liquid radwaste treatment plant as well as a few of solid radwaste treatment facilities were put into operation, other projects (New Safe Confinement, dry spent fuel storage etc.) are under way. Some progress has been reached also in development of national radwaste storages for low-level waste in the exclusion zone. However almost all projects were implemented with significant delays and budget overuses. In some cases the lack of coordination between different players still decreases efficiency and effectiveness of use of already commissioned facilities. The Chornobyl NPP decommissioning brings another issue to be considered in the long-term prospect, namely long-term availability of qualified staff to implement delayed decommissioning activities. Radiological survey has been completed to compare the current situation with previous forecasts and to update forecast of radiological situation in the exclusion zone. Taking into account its results, the proposed mid-term solution is to integrate all remediation and decommissioning activities under consolidated management within the inner part of the exclusion zone while establishing a national radiological natural sanctuary to proceed surveillance over cleaned peripherial areas of the exclusion zone, thus decreasing man-induced risks and allowing further developments. Pros and contras of such approach are discussed in the paper.

Speaker: Mr Vitalii Petruk (State Agency of Ukraine for the Exclusion Zone Management)

15:00 → 16:00 Session 1-1: Establishing National Policies and Strategies to Enable and Enhance D&ER 1

The purpose of this session is to demonstrate how national policies have been developed and are being implemented that enable and enhance D and ER and that are in line with widely accepted international principles; stimulate countries to develop national framework to enable D&ER; including skills development.

16:00 → 16:30 Coffee break

16:30 → 18:15 Session 1-2: Establishing National Policies and Strategies to Enable and Enhance D&ER 2

The purpose of this session is to demonstrate how national policies have been developed and are being implemented that enable and enhance D and ER and that are in line with widely accepted international principles; stimulate countries to develop national framework to enable D&ER; including skills development

16:30 The Current Situation of Off-site Clean-Up in Japan

The Current Situation of Off-site Clean-Up in Japan Kazumi Yoshikawa Director, International Cooperation Office for Decontamination Radioactive Materials, Environment Management Bureau Ministry of the Environment(MOE),Japan 1.Clean-up Activities After nuclear accident at TEPCO Fukushima Daiichi Nuclear Power Station, the Act on Special Measure was established in August 2011 to deal with environmental contamination by radioactive substances diffused. Under this Act, MOE has been implementing decontamination. To be more specific, the Act sets two types of areas for decontamination. The first category is the Special Decontamination Area (SDA) where the decontamination is conducted directly by Japanese Government. The second category is the Intensive Contamination Survey Area (ICSA), whose decontamination is conducted by the municipalities. The SDA is the area where the evacuation order was issued, where 11 municipalities within Fukushima Prefecture were designated by MOE. In this area, MOE established the decontamination plan and implemented the work. Concerning the SDA, the decontamination work is planned to be completed by March 2017. We will do our best for the decontamination work steadily, although we also have some tasks to obtain the consent from landowners & house owners for decontamination and secure the temporary storage sites. On the other hand, the ICSA is the area where the air dose rate is rather low, anticipated to exceed 0.23μSv per hour. This area was also designated by MOE, but designated municipalities set their own decontamination plans and implement them by technical and financial support from MOE. There are 94 out of 99 ICSA municipalities designated by the government and municipalities which planned the decontamination, half of them were completed with decontamination, and half of them are under on-going decontamination. ICSA within Fukushima Prefecture, about 90% of public facilities in the living area was completed. Outside Fukushima Prefecture, planned decontamination work was almost completed and coming close to the end.With the cooperation of relevant ministries, MOE will continue to provide technical and financial support in order to progress the decontamination work and speed-up the reconstruction of evacuated areas. 2.Interim Storage Facility and the Idea of Final Disposal As the Interim Storage Facility, it is estimated that generated soil from decontamination will be about 16-22 million m3 after the volume reduction. One of the biggest issues for the ISF is the transportation as well as the construction. We have to take all measures possible to transport a large quantity of decontamination soil safely and securely. For this reason, pilot transportation is implemented for about a year in order to confirm safety. According to the law, Japanese Government shall take necessary measures for the final disposal outside Fukushima Prefecture within 30 years after the start of the Interim Storage. To do this, 8 steps would be taken for the final disposal outside Fukushima Prefecture. As a first step, Japanese government will conduct research and develop technology of the volume reduction and recycling for final disposal. At the same time, we will share information with the public to build the public consensus for reuse of low radioactive materials and the final disposal outside Fukushima Prefecture. In order to proceed with the review of the final disposal site, we need to identify the characteristics, radioactive concentration and quantity of the soils generated by volume reduction technology. Reflecting this study, we also find out the final disposal method. Through such review, we will develop the strategy for volume reduction and recycling about the next 10 years, while securing a budget. 3.Public Communication and Cooperation with International Societies To share information with local community, MOE made easy-to understand pamphlets and other materials. Latest information is also uploaded on the MOE-Website. For international community, MOE has asked for advice and evaluation from IAEA and also held bilateral meetings with US,UK and France to exchange information for effective environmental remediation efforts. REFERENCE 1. Ministry of the Environment ,Japan(MOEJ),2015 “Progress on Off-site Clean-Up and Interim Storage in Japan” Page 1-56 2. Ministry of the Environment ,Japan(MOEJ),2015 “FY2014 Decontamination Report-Digest Version-“ Page1-78 3. Ministry of the Environment ,Japan(MOEJ),2015 “The Current Situation of Decontamination and Interim Storage” Page14-31(Japanese)

Speaker: Mr Kazumi YOSHIKAWA (Ministry of the Environment,Japan)

16:50 ESTABLISHING NATIONAL POLICIES AND STRATEGIES TO ENABLE AND ENHANCE DECOMMISSIONING

On March 11, 2011, the Fukushima Daiichi Nuclear Power Station (NPS) suffered a severe accident brought about by the tsunami caused by the Great East Japan Earthquake. The decommissioning of the Fukushima Daiichi NPS is a globally unprecedented challenge that involves a lot of difficulties. For this task to be implemented smoothly, it is not only essential to advance potential technologies that incorporate knowledge gathered from around the world. 1) Overview of Decommissioning Strategy The Government of Japan made “Mid-and-Long Term Roadmap towards the Decommissioning” and has updated it in 2013 and 2015. On the roadmap, there is 3 phases. Phase1 has completed in November 2013, with the start of removal of spent fuels from Unit4. We are now in the beginning of Phase2, which will be completed with the preparation for removal of fuel debris, which is planned on the end of 2021. In Phase3, actual decommissioning activities will be carried out in 30-40 years, which is likely very long and challenging period. 2) Main Progress on Decommissioning Work Removal of fuel from the spent fuel pool in Unit 4 was completed in December 2014, and the Unit is being maintained in a stable condition. As for Unit 3, large-size rubble has been removed out of the upper section of the reactor building. Presently, removal of roughly 20 tons of large-size rubble out of the spent fuel pool is underway by remote control. Cooling is continuing stably in all of the Units. 3) Main Progress on contaminated Water Management For management of contaminated water, the Government of Japan defined comprehensive countermeasures to manage contaminated water, based upon 3 major principles, (1) “Removing” the contamination source, (2) “Isolating” groundwater from the contamination source and (3) “Preventing leakage” of contaminated water. These countermeasures have been progressing according to the structured plan defined in the countermeasure, such as groundwater by-pass system, sub-drain system, impermeable walls with a soil freezing method, seaside impermeable walls, etc. 4) Cooperation with International Communities In order to complete this extremely difficult mission and to contribute to the global nuclear community, we are committed to disclose or share our new findings in our operations to the world. From this context, we cooperate closely with the IAEA and have requested peer-review mission in 2013 and 2015. Furthermore, on April 2016, we hold “The 1st International Forum on the Decommissioning of the Fukushima Daiichi NPS” at Iwaki-city, Fukushima.

Speaker: Mr Hirohide Hirai (Ministry of Economy, Trade and Industry (METI))

Tuesday, 24 May

09:00 → 11:00 Session 2 - 1: Regulatory Framework and Standards for D&ER 1

The purpose of this session is to discuss significance of regulatory requirements to implementation of D&ER programmes, including discussion of evolution of regulations and regulatory approaches

10:15 SUCCESSFUL IMPLEMENTATION OF WENRA SAFETY REFERENCE LEVELS: THE WENRA DECOMMISSIONING REPORT

Abstract: The Western European Nuclear Regulators Association (WENRA) was estab-lished in 1999. Its Working Group on Waste and Decommissioning (WGWD) has developed Safety Reference Levels (SRL) reports for Decommissioning[1], Storage[2] and Disposal[3] according to the original mandate. WENRA members have experienced a benchmarking process and established National Action Plans (NAP) for the modification of their national legal systems and practices according to benchmarking results. For the decommissioning and disposal reports the NAP have been implemented and results were approved in a follow up benchmarking exercise by WGWD. The whole process is explained for the decommis-sioning SRLs in this presentation. WGWD is currently working on developing a last SRLs report for waste processing which will complete the comprehensive description of the back-end of the nuclear fuel cycle. 1. INTRODUCTION The Working Group on Waste and Decommissioning (WGWD) comprises representatives from all 18 members . WGWD choose a holistic view to establish comprehensive SRL-sets for the thematic areas. In each report common safety areas such as safety management or safety verification are supplemented by report-specific safety areas, in case of the decom-missioning report Decommissioning Strategy and Planning and Conduct of Decommission-ing. Before finally publishing the SRLs WGWD had published the drafted texts, called for stakeholder comments and invited the responding stakeholders to a workshop for discussing their comments and –in some cases- optimizing the SRL-wording. 2. METHODS Each member country had to provide in a self-assessment table evidence from its regulatory system for fulfilment of all 81 SRLs . This information was subject to a panel benchmark-ing procedure within WGWD which worked in 4 subgroups for this purpose. Figure 1 shows the condensed result of this exercise; A-rating is equivalent to implementation of an SRL in the regulatory system, B-rating indicate rare cases of justified deviations and C-ratings in-dicate deficiencies. Figure 1. First benchmarking results by safety issues 3. RESULTS During the following two years WENRA-countries took efforts to improve their regulatory systems to –ideally- cover all identified deficiencies and reported on such activities to the working group. In a follow up “re-benchmarking” these reports of corrective actions have been affirmed by the WGWD. Results are documented in the country fact sheets of part III in the final report version 2.2. Country fact sheets contain a textual description of the cor-rective actions and a table including the new status and the citation of relevant regulatory texts. 4. CONCLUSIONS The most important benefit of the work in WENRA is learning from each other in • Understanding all facets of a given situation or problem • Identifying and comparing different ways of regulatory response • Mutual support and confirmation in identifying appropriate improvements Nothing can explain in a better way the effectiveness of WENRA work, but the fact that within only 3 years between the first benchmarking and the follow up the great majority of member countries could practically eliminate their C-ratings for the decommissioning SRLs or at least have prepared texts for corrective actions and are close to doing so. Another more general experience is understanding the variety of regulatory instruments even in Europe which include -in individual countries- acts, laws, ordinances, regulatory guides, regulatory orders, royal decrees or standardized license conditions. Finally it is to be highlighted that compared to the storage report which describes a static, steadily operated facility the decommissioning report describes a continuous process. It is one of the greatest challenges to properly address this difference but not forget about the common features in editing the SRL-sets for those two reports. 5. OUTLOOK WGWD is planning to establish a last set of SRLs –on waste processing- before end of 2016 and to carry out a similar benchmarking procedure as described also for the disposal and the processing report. From the experience of the first two reports it will be shortly before end of the decade that we might be in a position to report “mission completed”. In the meantime we will be happy to communicate our results to other organizations such as the IAEA, NEA or –our most direct counterpart- the ENISS. REFERENCES [1] WENRA, Report: Decommissioning Safety Reference Levels, V. 2.2, Apr. 2015, available on: www.wenra.org. [2] WENRA, Report: Waste and Spent Fuel Storage Safety Reference Levels, V. 2.2, Apr. 2014, available on : www.wenra.org. [3] WENRA, Report: Radioactive Waste Disposal Facilities Safety Reference Levels, Dec. 2014, available on: www.wenra.org

Speaker: Mr Stefan Theis (ENSI, Switzerland)

09:00 → 13:30 Session 2 - Poster

09:00 Challenges in Regulatory Oversight on Decommissioning for the First Time in Thailand

This paper covers various regulatory issues and challenges of decommissioning the first research reactor in Thailand. It will also be the first decommissioning of a nuclear facility in Thailand. The paper particularly emphasizes regulatory development of Thailand on decommissioning a research reactor. The regulatory issues on decommissioning and challenges for Thailand are identified. The brief survey of available decommissioning regulations from other IAEA member states is presented. Finally, the proposed decommissioning regulations of Thailand is summarized and discussed.

Speaker: Mr Chaiyod Soontrapa (Office of Atoms for Peace)

09:00 CROSS-CHECKING OF THE CUBAN EXPERIENCE ON THE DECOMMISSIONING OF MEDICAL, INDUSTRIAL AND RESEARCH FACILITIES AND THE IAEA REQUIREMENTS GSR-PART 6

In the last decades a wide range of nuclear and other facilities using radioactive material have been designed, constructed, operated and some of them decommissioned because they have reached their design lifetime. Decommissioning of facilities using radioactive material, is increasing worldwide, so, operators, regulators, stakeholders and the international community pay attention to this fact, in order to ensuring and maintaining safety of workers, the public and the environment during and after completion of decommissioning. Several facilities have been decommissioned in the last decade in Cuba; therefore, some valuable decommissioning experience has been obtained by the Regulatory Authority (RA). The legal framework for decommissioning has a significant impact on the planning for and implementation of decommissioning, so the Cuban Regulatory Authority pay special interest in a revision of it, with the objective to establishing an adequate legal and regulatory framework in line with the international safety standards and good practices. The objective of this paper is identifying if the safety requirements, recommendations, the international good practice and guidance are included or not, in our regulations and in the manner to consider the decommissioning phase nationally. Also this paper intends to share valuable lessons learned derived from decommissioning cases that have taken place in our country in order to improve the way it addresses this issue and prevent recurrence of unsuccessful experiences in the future

Speaker: Mr Ramón Hernández Álvarez (Centro Nacional de Seguridad Nuclear)

09:00 Destroyed fourth unit of Chernobyl nuclear power plant – from Shelter to New Safe Confinement. Safety regulation

SUSHKO Tamara, Radioactive Waste Management Safety Unit, Division on Decommissioning, State Nuclear Regulatory Inspectorate of Ukraine (SNRIU), 9/11 Arsenalna street, Kyiv, Ukraine E-mail address: sushko@hq.snrc.gov.ua KONDRATIEV Sergii, Decommissioning Unit, State Enterprise «State Scientific and Technical Center for Nuclear and Radiation Safety» (SSTC), 35/37 Vasylja Stusa street, Kyiv, Ukraine E-mail address: sn_kondratyev@sstc.kiev.ua The most severe accident in the world history of nuclear energy happened at the fourth unit of Chernobyl nuclear power plant in 1986. As a result of it the power unit was destroyed. A localizing building under ruined unit was erected in six months and equipped with specific systems for dust suppression, neutron absorbent solution spray, monitoring, etc. A great amount of nuclear and radioactive materials was buried inside. That facility is called the Shelter Object (hereinafter referred to as SO). The Shelter Implementation Plan [1] (hereinafter referred to as SIP) has been implemented by the international community since 1997 and includes both urgent measures on stabilization and safety upgrading and long-term measures aimed at transforming the facility into an ecologically safe system. Currently the urgent measures have been completed and the major SIP project - the installation of the New Safe Confinement (hereinafter referred to as NSC) above the SO - is actively implemented. In 1997 the Nuclear Regulatory Authority of Ukraine (currently the State Nuclear Regulatory Inspectorate of Ukraine – SNRIU) faced the challenge to provide adequate safety regulation of SIP implementation whereas no relevant experience existed in the world. This complicated task with the significant involvement of technical support organizations (hereinafter referred to as TSO) is being successfully carried out. One of key aspects of effective support rendered by TSOs to SNRIU is their comprehensive cooperation at the international level (between TSOs of Ukraine, Germany, France, the USA). The regulatory approaches were declared in the Statement of Policy for SIP Safety Regulation [2]. Such approaches foresee that SNRIU establish purposes, principles and criteria for the SIP activities which are based on provisions for the use of nuclear energy. During the development of SIP projects the Licensee is obliged to demonstrate, that safety goals are gradually achieved and principles and criteria for safety are met with the planned projects. The issue of applicability of different specific regulatory safety requirements is reasonable to solve in the course of the licensing process by development and implementation of particular projects. SNRIU also set forth the following three SIP fundamental safety principles [3]: 1) radiation safety and ALARA principle 2) application of proven technologies and advanced international experience and 3) introduction of quality management system by the Licensee. Based on these principles, the safety cornerstones and the guidance for application of these principles were developed for the SIP safety regulation [4]. The safety cornerstones: - SO structural integrity, - accident prevention, - emergency preparedness and mitigation of accidents consequences, - nuclear safety (prevention of criticality), - radiation protection of personnel, the public and the environment, - radioactive waste management, - quality management and safety culture. Based on the above approaches, the Licensee implements specific projects on transformation SO into an ecologically safe system and demonstrates to the SNRIU and other regulatory authorities (RA) that safety goals are step-by-step achieved and safety principles and criteria as well as technical requirements on safety are properly applied. Successful implementation of SIP requires systematic constructive dialogue within a well-established licensing process between the Licensee/Contractors, on the one hand, and SNRIU/TSO’s, on the other hand. Such licensing process was established at the beginning of SIP implementation and approved by all the parties involved. As experience was gained, the licensing process constantly improved and became more detailed. Such step-by-step licensing process with a constructive day-to-day dialogue has minimized the risks of SIP projects and ensured optimization of SIP designs in terms of safety goals, minimization of personnel exposure, etc. A number of SIP early projects were approved by the SNRIU and progressed from the pre-design studies to detailed designs. They were essentially optimized, for example, the projects of SO structure stabilization, the integrated monitoring system and the new safe confinement. In the process of implementation of these projects at SO, the organization of construction and assembly work was also optimized, which reduced actual personnel exposure as compared to design-basis estimates. REFERENCES [1] TACIS DG IA, EUROPEAN COMMISSION, USA MINISTRY of ENERGY, Shelter Implementation Plan, 1997 [2] NUCLEAR REGULATORY ADMINISTRATION of UKRAINE, Statement of Policy for nuclear and radiation safety regulation for Chornobyl NPP Shelter Object, 1998 [3] STATE NUCLEAR REGULATORY COMMITTEE of UKRAINE, “Fundamental Principles of safety activities within the Shelter Implementation Plan”, 2005 [4] STATE NUCLEAR REGULATORY COMMITTEE of UKRAINE, “Guidance on the application of the safety principles during regulatory activities within the Shelter Implementation Plan”

Speakers: Mr Sergii Kondratiev (State Scientific and Technical Center for Nuclear and Radiation Safety), Mrs Tamara Sushko (State Nuclear Regulatory Inspectorate of Ukraine)

09:00 IMPROVEMENT OF THE REGULATORY FRAMEWORK IN THE FIELD OF RADIOACTIVE WASTE AND DECOMMISSIONING

Abstract: The paper describes the work done in order to improve the regulatory framework in the field of radioactive waste and decommissioning of nuclear and radiological facilities. The National Commission for Nuclear Activity Control, as the nuclear regulatory authority of Romania, has improved the regulatory framework developing the safety and licensing requirements in the field of predisposal and disposal of radioactive waste and decommissioning of nuclear and radiological facilities. The paper describes the content of predisposal management and decommissioning regulations. 1. INTRODUCTION In order to meet the requirements of COUNCIL DIRECTIVE 2011/70/EURATOM of 19 July 2011 establishing a Community framework for the responsible and safe management of spent fuel and radioactive waste [1], the regulatory authority of Romania developed new regulations and revised the existing regulations in the field of predisposal and disposal of radioactive waste, and decommissioning of nuclear and radiological facilities. The improvement of regulatory framework has been done in the framework of the Project “Regional Excellence Project on Regulatory Capacity Building in Nuclear and Radiological Safety, Emergency Preparedness and Response in Romania “. The objective of the project is to enhance the capabilities of the Romanian nuclear regulatory authority CNCAN in eight specific functional areas of work through exchange of experiences, best practices, and capacity building with the Norwegian Radiation Protection Authority and the International Atomic Energy Agency. Main activities of the project are summarized under the following subprojects: CNCAN1 - Enhancement of CNCAN capabilities for safety analysis; CNCAN2 - Enhancement of CNCAN capabilities for integrated management systems and knowledge management; CNCAN3 - Enhancement of CNCAN capabilities for inspections; CNCAN4 - Enhancement of CNCAN capabilities for safety and security of transport and transit of radioactive and nuclear materials on the Romanian Territory; CNCAN5 - Enhancement of CNCAN capabilities for emergency preparedness and response; CNCAN6 - Enhancement of CNCAN capabilities for ionizing radiation sources control; CNCAN7 - Enhancement of CNCAN capabilities for radioactive waste, spent nuclear fuel management, and decommissioning activities; CNCAN8 -Enhancement of CNCAN capabilities for safeguards. Total project budget is EUR 4,215,098 consisting of 85% allocated from Norway Grants and 15% to be provided in cash by the Romanian national co-financing. The Project duration is 31 months in the period from October 2013 until April 2016. 2. DESCRIPTION OF THE SAFETY REGULATION ON THE PREDISPOSAL MANAGEMENT OF RADIOACTIVE WASTE The regulation is based on the IAEA recommendations provided in the General Safety Requirements Part 5 Predisposal of Radioactive Waste [2], as well as on the Safety Reference Levels developed by the Western European Nuclear Regulators Association (WENRA) on the storage of radioactive waste and spent nuclear fuel [3]. The regulation contains specific requirements for each step of the predisposal management covers the control of generation, characterization and classification of radioactive waste, waste acceptance criteria, collection, segregation, treatment, conditioning and storage of radioactive waste and disused radioactive sources. The chapter on requirements for the development of predisposal radioactive waste facilities details the safety requirements for siting, design, construction, commissioning, operation and permanent shut down of the facilities. The regulation introduces the concepts of safety case, safety assessment and periodic safety review. 3. DESCRIPTION OF THE SAFETY REGULATION ON THE DECOMMISSIONING OF NUCLEAR AND RADIOLOGICAL FACILITIES The safety and licensing requirements on decommissioning cover both nuclear and radiological facilities. The regulation is based on the IAEA recommendations provided in General Safety Requirements Part 6 Decommissioning of Facilities [4], as well as on the Safety Reference Levels developed by WENRA on decommissioning [5]. The regulation defines the end state criteria, requirements for decommissioning strategies, planning of decommissioning activities, as well as the transition from operation to decommissioning phase and conducting of decommissioning actions. The regulation introduces the concepts of safety case and safety assessment, their contents being provided in the Annexes to the regulation. The requirements for final radiological verification are also provided. The content of the final radiological survey report as well as the final decommissioning report are provided. 4. CONCLUSIONS The safety requirements provided in the regulations on the predisposal management of radioactive waste as well as on the decommissioning of nuclear and radiological facilities are in line with IAEA recommendations and meet the requirements of the applicable European Council Directives. REFERENCES [1] COUNCIL DIRECTIVE 2011/70/EURATOM of 19 July 2011 establishing a Community framework for the responsible and safe management of spent fuel and radioactive waste, Official Journal of the European Union, L 199/48. [2] IAEA GSR Part 5 Predisposal of Radioactive Waste (2009). [3] WENRA - Waste and Spent Fuel Storage Safety Reference Levels Report (2014). [4] IAEA GSR Part 6 Decommissioning of Facilities (2014). [5] WENRA - Decommissioning Safety Reference Levels Report (2015).

Speaker: Dr Daniela Maria Dogaru (National Commission for Nuclear Activities Control)

09:00 Introduction on Revision of Nuclear Safety Legislation Related to Decommissioning of Nuclear Facilities in Korea

There are 25 units of nuclear power reactors in operation and 3 units of nuclear power reactors under construction in Korea as of October 31st 2015. However, there is no permanently shutdown nuclear power reactor and decommissioned or under decommissioning nuclear power reactor. There are only 2 research reactors being decommissioned since 1997. It is realized that improvement of the regulatory framework for decommissioning of nuclear facilities has been emphasized constantly from the point of view of IAEA’s safety standards. IAEA published the safety requirement on decommissioning of facilities on July 2014; its title is the Safe Decommissioning of Facilities, General Safety Requirement Part 6. According to follow up action on the result of IAEA’s Integrated Regulatory Review Service (IRRS) mission to Korea in 2011, regulatory framework for decommissioning of nuclear facilities in Korea was revised through comparing to IAEA safety standards. It was identified that items should be revised to improve the regulatory framework for decommissioning. Those are as follows: absence of legal definition of decommissioning, incomplete procedure for safety regulation after permanent shutdown, undetailed acceptance criteria for decommissioning plan, incomplete requirements for early preparing and periodic update of decommissioning plan, undetailed requirements on standard format and contents for decommissioning plan, and incomplete radiological criteria on site and building reuse after completion of decommissioning. Nuclear Safety Act related to decommissioning of nuclear facilities was revised and promulgated on 21st July 2015. As the lower statute of Nuclear Safety Act, Enforcement Decree of the Nuclear Safety Act and Enforcement Regulation of the Nuclear Safety Act were also revised and promulgated on 21st July 2015. In this paper, related to decommissioning of nuclear facilities such as nuclear power reactor, research or training reactor, and nuclear fuel cycle facility, it was introduced the main changes of the amended and promulgated Nuclear Safety Act on July 2015. It was also mentioned about the current issue in accordance with its implementation. Main contents of revised Nuclear Safety Act are that decommissioning plan should be submitted for nuclear installations to be constructed and operated, and this plan should be updated periodically. In addition, 3 years of grace period was set to submit preliminary decommissioning plan for the facility which has already been approved prior to July 2018. Preliminary decommissioning plan should be updated every 10 years, and regulatory body should review this document. According to the revised Nuclear Safety Act, in the case of a nuclear power reactor after the approval of the change on operating license for permanent shutdown, should submit final decommissioning plan within 5 years, and be approved by the regulatory body. On July 2015, Kori Unit 1 was determined not to apply its 2nd continued operation, and will expire in 2017. Kori Unit 1 will be the first case for the submission of final decommissioning plan.

Speaker: Dr JungJoon Lee (Korea Institute of Nuclear Safety)

09:00 Legal steps for decommissioning and remediation process of legacy uranium mining sites

The safe deployment of nuclear activities in Romania is provided by Law 111/1996, republished and completed by Law 193/2003 and Law 378/2013. The competent national autority in the nuclear fiel, wich has responsabilities of regulation, autorization and control as stipulated in this Law, is the National Commission for nuclear Activities Control (CNCAN). According art. 2b) provisions of the nuclear Law shall apply to design, possession, siting, construction-assembly, commissioning, operation, conservation and decomissioning of the mining and milling facilities for uranium and thorium ores and of the waste management facilities of the waste resulted from the mining and milling of uranium and thorium ores. The current regulatory frameork in force for mining and milling is: 1. Law no. 111/1996 on the safe deployment, regulation, authorization and control of nuclear activities, republished; 2. Radiological Safety Norms on Decommissioning of Uranium and/or Thorium Mining and Milling Facilities with 3 Gidelines: a) Guide on the recommended parameters for the estimation of the effective doses on all relevant ways of exposure at the usage for different purposes of contaminated sites, dumps and tailing pound resulted from uranium mining and milling; b) Guidelines on criteria for release from CNCAN regulatory control in order to use for other purposes of the buildings, dumps and contaminated areas by uranium mining and milling activities; c) Guidelines on technical requirements for design, sitting, construction , operation and decommissioning of storage areas of uranium and thorium ores and radioactive waste from uranium milling. 3. Radiological Safety Norms on the Management of the Radioactive Waste resulted from Fuel Cycle Facilities. The provisions of this norms and guidelines provide elementary safety standards for the health protection of both workers and the general public against the dangers arising from ionizing radiation. In the same vein the specific regulations and guidelines contain detailed provisions on: the content of radiation protection program, conditions for release in the environment of liquid and gaseous effluents, monitoring and surveillance programs of the radioactivity of the environment factors, long term stability of damps and tailing ponds, decommissioning of uranium production facilities, responsibilities of the license holders, obligations of the license holders regarding non-radiological risks generated by radioactive waste arisen from uranium mining and milling. According to norms in force a proposal for the derived effluent emission limits must be initiated by the holder of license and approved by the regulatory authority. These limits must ensure full compliance with the dose constraints for member of the public. For example, for the closure of a uranium mine, the regulatory authority imposes a constraint of 0.3 mSv/year (or less if the collective dose is too high) for each individual within the critical group. Also the norms stipulate the obligation of the holders to assure treatment of mine waters long time after closure in order to reduce the uranium and radium contents at accepted levels for general public. Total daily volume of the mine waters discharged in environment is about 10,000 m3/day containing up to 3mgU/l and 0.2 Bq Ra/l. After treatment (using ion exchange technology), their uranium average content is about 0.2 mg U/l. The regulations for uranium mining and milling specify the radiological protection principles required for the use or release of contaminated areas, buildings, metallic scrap, sites and dumps from uranium facilities and for tailing pound closing. For the unrestricted use of the reusable equipment and devises a contamination limit of 0.05 Bq/cm2 total alpha activity was established. The areas contaminated by the uranium mining can be released for unrestricted use if after remediation the activity of Radium-226 in dry soil is less than 0.2 Bq/g and the dose rate of doze measured at 1 m on top surface is less than 30 nSv/hou. With regards to uranium mining and milling activities, the Romanian legislation takes into account three objectives: - to ensure that the workers, the public and the environment are protected against the radiological hazard resulting from the exploitation of the uranium mining and milling industries; - to provide protection during the period of exploitation and after the closure of uranium mines or milling facilities; - to ensure that wastes (dumps) resulting from the operating of uranium mines and tailing ponds resulting of uranium ore milling are trated as radioactive waste.

Speaker: Mrs LORENA POP (NATIONAL COMMISSION FOR NUCLEAR ACTIVITIES CONTROL)

09:00 Legalization of In-Situ Remediation Practice of Contaminated Drain Pits by TE-NORM close to El-WAHAT Libyan oil fields

ABSTRACT Low Specific Radioactivity Scales (LSRS) that associated with Libyan oil and gas production facilities are of natural origin. Its recognition, existence and significance were studied by oil and gas industries in Libya since 1997. Surface disposal of radioactive sludge/scale, and produced water (as practiced in the past) might lead to ground and surface water contamination. Because TE-NORM contaminated wastes in oil and gas production operations were not properly handled. Thus disposal of generated wastes that resulted in environmental contamination in and around production facilities require promulgated practices. Thus in this paper the author trying to stress on the lack of promulgated legislative framework and its impact on safety of oil and gas worker in Libya and, to highlight the actions that has been carried out to come over the problem. INTRODUCTION In most countries promulgated rules and regulations governing and controlling the utilization of sources of ionizing radiation were intended to be applicable to products of nuclear industry. By the time of oil discovery and particularly in middle of 20th century many national governments have been reluctant to get involved in regulations of controlling radioactive materials of natural and/or environmental origin. Lately and after many researches it becomes clear that, the exposure of human beings to ionizing radiations from NORM/TE-NORM sources is of great concern. Such concern was so important particularly, to national authorities dealing with public health and protection of radiation workers. Although TE-NORM is generated from natural sources and become in existence due to human activities. Regulations concerning both safety and waste handling of TE-NORM have generally been derived from different safety standard that were put for handling, management, control, and disposal, of man-made radioactive sources. Thus many national and international associations and commission contributed to total integrated radiation dose to human from all radioactive sources, among them: ICRP, EPAs, UNSCAR, CRCPD, IAEA, HPS, NRCs, IUR, NRPB, EURATOM, NNRs, and IRPA. METHODOLOGY In-Situ Remediation procedures that require legalization was conducted by Wintershall-Libya to clean and remediate NORM contaminated area measuring about 35,000 square meter, and comprises of two drain pits and a slim oil pit that were operated between 1977 to 2004 for discharge of produced water in the middle of Libyan Sahara near EL-WAHAT Oasis. The remediation procedure conducted in Four phases they were: Phase one: Preparatory measures to allow for remediation. Phase Two: Monitoring and contouring of site to be remediated to allocate hot spots and to predict level of contamination. Phase Three: Construction of the covers that needed for remediation process. Phase Four: Closing measures, that include verification of remediation success. RESULTS: In-situ approaches as criterion for remediation of NORM/TE-NORM contaminated pits that performed by Wintershall allow for 60% reduction in average gamma dose rate on the disposal site, taking into consideration levels of doses defined in NOC “NORM Management Manual”. The outcome of such procedure produced water disposed by salt water disposal system after separation and pumping it back in to reservoir. Additionally this In-Situ process ensures long-term stability of remediated site against erosion process due to construction of radiation shielding cover on the refilled pit. Conclusion Although the conducted criterion for remediation of contaminated pits by NORM was not approved by General Environmental Authorities of Libya. It is evident that the huge existing quantity of TE-NORM waste generated annually are contributing to the growing radiation exposure of workers and members of public. Lack of regulatory infrastructure impedes development of national standards for the control of TE-NORM outside of the coverage of the Atomic Energy Commission. In general the applied procedure “In-Situ Remediation Practice of Contaminated Drain Pits by TE-NORM close to El-WAHAT Libyan oil fields” was successful, applicable, and might be effective. Further investigations are still required for making such procedure generalized. However, when a country need to establish a promulgated regulatory framework for controlling, management, and disposal of TE-NORM four important radiation protection issues need to be considered: (i) technical enhancement operations, (ii) TE-NORM deposition, (iii) remediation procedures, and (iv) exemption levels. REFERENCES [1] AL-MASRI, M. S., ABA, A., Distribution of Scales Containing NORM in Different Oil Fields Equipment. Appli. Radia. Isot. May (2005), 55:1-7. [2] FAWARIS , B. H., .and AHMAD, A. SAAD , Natural Radioactive Scales in oil and Gas Fields of Jamahiriya: A Challenge to Radiation Safety of Workers. Presented in the 7th Arab Conference on the Peaceful uses of Atomic Energy, Sanaa - Yamen 4-8 Dec., (2006).

Speaker: Bulgasem El-fawaris (LAEE-Libya)

09:00 Lessons learned from the first-time application of current decommissioning regulations in Sweden, and possible improvements to future regulatory supervision

Economic considerations have resulted in decisions by nuclear power plant owners to advance the dates for the permanent shut down of four of Sweden’s ten operating nuclear power reactors. These closures are to occur within the next five years and the decommissioning of the reactors is planned with very short transition and care and maintenance periods. This accelerated schedule is likely to put stress on a number of the arrangements for the waste management system in Sweden, including the national financing arrangements for waste and decommissioning and the supervisory work of the national nuclear and radiation safety regulator. In order to meet the latter challenge, the Swedish Radiation Safety Authority (SSM) will have to apply lessons learned from regulatory supervision of decommissioning projects to-date. The presentation will highlight lessons learned from the first-time application of current regulations during regulatory supervision of ongoing Swedish decommissioning projects. It will identify possible improvements to coming regulatory supervision as well and how these could be addressed in the review of SSM’s current decommissioning regulations. Currently there are three decommissioning projects being undertaken in Sweden: two boiling water reactors at the Barsebäck nuclear power plant; two material test reactors in Studsvik; and the uranium mining and milling facilities in Ranstad. Barsebäck Unit 1 was shut down in 1999, followed by Unit 2 in 2005. Since then these units are in a care and maintenance phase. In 2016 the reactors internals are planned to be dismantled and temporarily stored in a new facility on the site. The reactors in Studsvik (one tank type and one mobile pool type) were permanently shut down in 2005. The reactors were dismantled in 2015 and preparations for decommissioning of the biological shield and the remainder of the facility are ongoing. The uranium mining and milling facilities in Ranstad were constructed and operated in the 1960s. The uranium mine and mill tailings deposits were restored and covered in the 1990s. Currently, decommissioning of the remaining facility at Ranstad is proceeding and is planned to be concluded in 2017.

Speakers: Mr Martin Amft (Swedish Radiation Safety Authority (SSM)), Mr Mathias Leisvik (Swedish Radiation Safety Authority (SSM))

09:00 Modification of the Hungarian regulatory system related to the oversight transfer

The Act VII of 2015, which is made for the capacity expansion of Paks nuclear power plant in order to establish integrated atomic energy oversight, has given the scope of duties of radiation protection to the atomic energy oversight organization from January 1, 2016. The purpose of the integration is to make the nuclear safety, radiation protection, physical protection to belong to the same authority. The National Health Office will remain the competent authority of the radiation health questions. On the basis of the modification of the Act on Atomic Energy, the licensing and inspection of siting, construction, operation, modification and closure of radioactive waste repositories (Radioactive Waste Treatment and Disposal Facility and National Radioactive Waste Repository) belongs to Hungarian Atomic Energy Authority (HAEA), as the atomic energy oversight authority since June 30, 2014. HAEA performed the necessary actions to take over the regulatory processes under way and in order to assess and enhance the present safety of waste repositories, verify the compliance of the local processes with the related legislative prescriptions. Considering the decommissioning and the environmental remediation programme, the radiation protection and the radioactive waste management are very important issues. Because of the changing of the Hungarian regulatory body the presentation will show the legislative and regulatory system in Hungary and some aspects of the Hungarian experience in aspects of the scope of duties. The nuclear energy has a significant role in the Hungarian electricity system. In a few words we present the Hungarian nuclear programme in radioactive waste, spent fuel and decommissioning related issues.

Speaker: Mr Zsolt Sebestyén (Hungarian Atomic Energy Authority)

09:00 RECENT PROGRESS IN THE STUDIES OF FORMAT AND CONTENT OF SAFETY ANALYSIS REPORTS ON FACILITATING DECOMMISSIONING FOR CHINA NPP

Abstract:Nuclear power plants usually have a typical design life of 40 years which can be extended up to 60 years. At the end of their operating life-times, they need to be decommissioned to ensure the safety. The design and operational management of early nuclear power plants were given inadequate consideration for the decommissioning resulted in difficulty in dismantling reactor, large amounts of waste, and high costs when implementing decommissioning. Decommissioning has become an important issue that has hindered the development of nuclear power. Safety analysis report is a major technical document reviewed by regulatory authority, and therefore should include the relevant content on facilitating decommissioning. China National Nuclear Safety Administration has asked the operator to prepare an independent chapter (Chapter20) for describing design features and operational measures to facilitate decommissioning in safety analysis report for nuclear power plants in 2013. This paper introduces the format and content of the newly developed Chapter 20 of the safety analysis reports for academic exchanges. REFERENCES [1]Law of the People's Republic of China on Prevention and Control of Radioactive Pollution(2003). [2]Ministry of Environmental Protection/National Nuclear Safety Administration of the People’s Republic of China, Regulations of the People’s Republic of China on the Safety Control of Civilian Nuclear Installations, Chinese HAF Series No.001 (1986). [3]Ministry of Environmental Protection/National Nuclear Safety Administration of the People’s Republic of China, Detailed Rules (I) of the People’s Republic of China on Regulating Civil Nuclear Facility Safety-Application and Granting of NPP Safety Licenses, Chinese HAF Series No.001/01(1993). [4]Ministry of Environmental Protection/National Nuclear Safety Administration of the People’s Republic of China, Regulations on the Safety of NPP Design, Chinese HAF Series No.102(2004). [5]Ministry of Environmental Protection/National Nuclear Safety Administration of the People’s Republic of China, Design of NPP Reactor Containment System, Chinese HAD Series No.102/06 (1990). [6]Ministry of Environmental Protection/National Nuclear Safety Administration of the People’s Republic of China, Design of NPP Reactor Core Safety, Chinese HAD Series No. 102/07 (1989). [7]Ministry of Environmental Protection/National Nuclear Safety Administration of the People’s Republic of China, NPP Reactor Cooling System and Related Systems, Chinese HAD Series No.102/08 (1989). [8]Ministry of Environmental Protection/National Nuclear Safety Administration of the People’s Republic of China, Fire Protection at NPPs , Chinese HAD Series No.102/11(1996). [9]Ministry of Environmental Protection/National Nuclear Safety Administration of the People’s Republic of China, Design of Loading/Unloading and Storage System at Nuclear Power Plant, Chinese HAD Series No.102/15 (2007). [10]Ministry of Environmental Protection/National Nuclear Safety Administration of the People’s Republic of China, Safety Analysis and Verification for Nuclear Power Plant, Chinese HAD Series No.102/17 (2006). [11]Ministry of Environmental Protection/National Nuclear Safety Administration of the People’s Republic of China, Regulations on Radioactive Waste Safety, Chinese HAF Series No.401 (1997). [12]Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources, Chinese national standards: GB18871 (2002). [13]Regulations on Radiation Protection for NPPs, Chinese national standards: GB6249 (2011). [14]Requirements for Nuclear Facility Decommissioning, Chinese national standards: GB/T19597(2004). [15]Regulations on Radiation Protection for Decommissioning NPPs and Large Sized Reactors, Chinese national standards:GB11850(1989). [16]Regulations on Environmental Management Techniques for Decommissioning Reactor, Chinese national standards: GB/T14588 (2009). [17]Regulations on Radioactive Waste Management, Chinese national standards: GB14500 (2002). [18]Regulations on Radiation Protection for Decommissioning Nuclear Fuel Reprocessing Plant, Chinese Nuclear Industry Standards:EJ588(1991). [19]Safety of Nuclear Power Plants: Design, IAEA Safety Standards Series No. SSR-2/1(2012). [20]Decommissioning of Nuclear Power Plants and Research Reactors Safety Guide, IAEA Safety Standards Series No.WS-G-2.1(1999). [21]Safety Assessment for the Decommissioning of Facilities Using Radioactive Material, IAEA Safety Standards Series No.WS-G-5.2(2009). [22]Experiences and Lessons Learned Worldwide in the Cleanup and Decommissioning of Nuclear Facilities in the Aftermath of Accidents, IAEA Nuclear Energy Series No.NW-T-2.7(2014). [23]Safety Assessment for Decommissioning, IAEA Safety Reports Series No.77(2013). [24]Design Lessons Drawn from the Decommissioning of Nuclear Facilities, IAEA TECDOC No. 1657(2011). [25]Long Term Preservation of Information for Decommissioning Projects, IAEA Technical Reports Series No.467(2008). [26]Disposal Aspects of Low and Intermediate Level Decommissioning Waste, IAEA TECDOC No.1572(2008).

Speaker: Mr Xiaolong LI (Nuclear and Radiation Safety Center,Ministry of Environmental Protection of the People's Republic of China)

09:00 REGULATION OF DECOMMISSIONING ACTIVITIES IN THE SLOVAK REPUBLIC

The paper describes the situation and activities in the field of decommissioning in the Slovak Republic. Systematic approach of decommissioning in Slovakia can be date back to the 1990's when the Project of “Putting the NPP A-1 into radiation safe status" without spent fuel and without uncontrollable release of radioactivity into environment was commissioned. At the end of 1990's the update of obsolete “atomic act” came into force when previous act dates back to 1984. On the base of “atomic act” there were specified conditions for approval of decommissioning and the licence for 1st decommissioning stage was issued by regulatory body. The paper summarizes the lessons learned from the regulation of decommissioning activities over this period. The actual decommissioning program consists with three ongoing decommissioning projects. Decommissioning of a HWGCR type of reactor KS- 150 (NPP A-1) shutdown in 1977 after primary coolant system integrity accident with local melting of the fuel, decommissioning of two VVER 440 reactors of V-230 (NPP V-1) shut down in accordance with the Governmental resolution in 2006 and 2008 respectively and decommissioning of two experimental nuclear facilities – a bituminisation line and an incinerator for RAW owned by VUJE Plc installed in the premises of NPP A1 in Jaslovske Bohunice. Meanwhile the incineration facility was released from nuclear regulatory control last year. Decommissioning of NPPs is operated by JAVYS, Plc. Currently 2nd stage of the A1 NPP decommissioning project is in progress. The activities concentrate on decommissioning of external structures of the A1 NPP, on the problems related to the management of contaminated soil and on the management of RAW from the main production compound of the A1 NPP. The completion of 2nd stage of the decommissioning is planned for the end of 2016. Actually there is in preparation in 3rd and 4th decommissioning stages that is going to be licensed together. The most challenging for the project preparation will be the last decommissioning stage that ends in 2033. The objective of this fifth stage will be mainly focused on dismantling of reactor, steam generators and main transport technological part of primary circuit. During this phase, the cardinal part of RAW which will not be accepted for near surface disposal will be originated. The object of the 2nd (last) stage of the V1 NPP (2015–2025) is the decommissioning of structures in the main production compound, the buildings of auxiliary operations and the remaining auxiliary buildings, which will not have been decommissioned during the 1st stage of decommissioning, or which will have been only partially decommissioned, and the final site management. The most important activities during the 2nd stage of decommissioning will be the dismantling of the reactors and of the primary circuit equipment. Other activities, such as the dismantling of other general equipment in the controlled zone and outside the controlled zone, the decontamination and radiation inspection of structure surfaces, the demolition and the management of the site are standard activities, for which there is enough experience to go on, from other decommissioning projects. Decommissioning activities are regulated by Nuclear Regulatory Authority of SR (UJD SR) as the central state authority for nuclear safety supervision. The legal basis for decommissioning is included in the actual Act No 541/2004 Coll on peaceful use of nuclear energy (atomic act). The act was prepared by UJD SR and it is based on internationally agreed principles. One of the basic roles of UJD SR is creation of the legislative environment. This area has been recently dominated by works on several amendments of the atomic act due to consistent transpositions of individual Council Directives among others the Council Directive 2011/70/Euratom establishing a Community framework for the responsible and safe management of spent fuel and radioactive waste. The effectiveness of the Slovak regulatory framework for nuclear safety that is under the competency of UJD SR was subject of the IAEA mission focusing on strengthening and increasing the effectiveness of national regulators (IRRS). Particular attention was paid to review of decommissioning and radioactive waste management framework against the safety standards of the IAEA, as well as against the international criterion for safety. The paper presents challenges in the field of planned updating of decommissioning framework that can be summarised as follows: Phase out of stage approach, intention is to issue one overarching decommissioning licence supported by individually licensed decommissioning projects; State responsibility for decommissioning by establishing of a legal entity which has been founded, or authorised by the Ministry of the Economy, Updating of atomic act, new version that comprises a transposition of Council Directive 2014/87/Euratom and 2013/59/Euratom, implementation of IRRS finding, establishing conditions for transmitting of the NPP to the legal entity for the decommissioning purpose.

Speaker: Mr Miroslav Drahos (Slovak)

09:00 REGULATORY FRAMEWORK AND STANDARDS FOR DECOMMISSIONING AND ENVIRONMENTAL REMEDIATION, MONGOLIA

Mongolia is the country that is under second generation of activity of radioactive minerals’ exploration. Three main areas are covered by intensive uranium exploration activity. They are respectively Chuluut, Dulaan Uul (including uranium deposits Haraat, Hairhan, Zuuvch Ovoo etc.) and Mardai area, which are recently covered by tenements owned by well known companies. Most of these areas are under naturally normal condition except Mardai area. Mardai area, covers 50x25km land, included the biggest uranium deposits in Mongolia, namely Dornod, Gurvanbulag, Mardai, Nemer and other satellite small deposits, Har, Havar, Ilreh, Tsever etc. All of these uranium deposits are revealed by a great scope of exploration work completed by Soviet Union during 1980s. To explore in detail and to prepare the selected deposits for mining, they have excavated/exploited and dumped couple of thousand tons of waste rocks and piled several hundred cubic meters of high and low grade ores on the surface for loading into the trains and for heap leach testing. Also, different sizes of diggings stay in adjacent. Following the reforming in the Soviet Union, beginning of 1990s, all of the experts and workers had been backed up and left the waste rock dumps and approximately 0.16-0.30% U graded mineralized rock piles where they were on the surface. They are still appearing everywhere in the area, for example by the rail beds, by the shafts, at the mine sites for more than 20 years without controlling. If the deposits are not going to be mined in near term, the waste rock dumps and ore piles should be managed for each of deposits according to the“Management of NORM residues”by IAEA (iaea-tecdoc-1712), Basic Safety Standards and other IAEA publications. Regulatory bodies of Mongolia are working on national policies and regulations some of which are related to this title.

Speaker: Mr Bat-Erdene Baatar (Mongolia)

09:00 REGULATORY FRAMEWORK AND STANDARDS FOR DECOMMISSIONING AND ENVIRONMENTAL REMEDIATION; A CASE STUDY OF GHANA.

Ghana’s quest for expanding the horizon of her peaceful application of nuclear science and technology and in fulfilment of her international obligations especially the International Atomic Energy Agency’s Fundamental Safety Principles and General Safety Requirements has led to the passage of Nuclear Regulatory Authority Bill 2014 by the Parliament of Ghana in August 2015 thereby paving the way for the deployment of Nuclear Power Plants into the national energy mix. Ghana currently runs a 30kW research reactor which is yet to be decommissioned. The country therefore does not have legacies from the development of nuclear energy, including the dismantling of redundant research and fuel cycle facilities, research reactors and nuclear power plants, and the remediation of former nuclear sites and those affected by past uranium mining and processing operations or other activities involving the use of naturally occurring radioactive materials (NORMs) or by major nuclear or radiological accidents. The Bill encompasses modern best practices in the nuclear industry with respect to regulatory framework and standards for decommissioning and environmental remediation. This paper is therefore meant to illuminate this vital aspect of the nuclear industry in Ghana. These include decommissioning plan, financing of decommissioning, obligations of authorized person for decommissioning and decommissioning of nuclear facilities. The actualization of this bill will ensure that Ghana’s current and future nuclear energy activities will not pose unbearable burden on generations yet unborn.

Speaker: Mr Gustav Gbeddy (Ghana Atomic Energy Commission)

09:00 Remediation of the Russian Nuclear Legacy Sites. Relevant Issues of Regulatory Supervision.

SHANDALA(1), Natalya, KISELEV(1), Sergey, SEMENOVA(1), Mariya, SNEVE(2), Malgorzata, SMITH(3), Graham (1)- Public radiation protection department, SRC FMBC, Zhivopisnaya str., 46, Moscow, Russia (2)- Norwegian Radiation Protection Authority, Postboks 55, 1332 Østerås, Norway (3)GMS Abingdon Ltd, Tamarisk, Radley Road, Abingdon, Oxfordshire, OX14 3PP, UK E-mail address: sergbio@gmail.com Abstract: As part of the national policy of the Russian Federation in management of spent fuel and radioactive waste in the Russia, the Federal Targeted Program on Nuclear and Radiation Safety is approved. This Program provides a landmark reduction in amount of nuclear and uranium legacy. Start of this work (regulatory issues) at the legacy sites was given 10 years ago in the framework of the Russian-Norwegian cooperation at the nuclear legacy sites(NLS) in the Northwest Russia, as well as in the implementation of the Federal Targeted Program "Decommission and Dismantlement of arms and military equipment" (2005) (legacy sites in the Far East Russia). The accumulated experience in enhancing the safety culture during remediation of former naval shore technical bases in the North West and Far East Russian regions is discussed in this paper. 1. INTRODUCTION In the 1960s, in the Northwest and Far East regions of Russia, shore technical bases of the Navy were established. After the expiration of the project resources, the infrastructure of facilities degraded, resulting in a serious potential threat of radioactive contamination of the environment. According to the decree of the Russian Government these bases have been transferred under the management of the State Corporation "Rosatom" for the purpose of their environmental remediation. FMBA is the official regulatory body at the Rosatom’s facilities[1]. It is responsible for independent control and supervision of radiation safety. This paper is focused on the main results of scientific and practical activities in regulatory supervision during remediation of former shore technical bases of nuclear submarines of the Russian Navy. 2. METHODS In order to identify the most important issues requiring supervision and enhanced regulation during remediation of former technical bases, the initial radiological threats were assessed. The results of radiological threat assessment at the initial stage of the STS remediation, focused the regulatory activities in the following areas: RP of the workers, public&environment RP, emergency response and preparedness[2]. 3.RESULTS Protection of workers. The AndreevaPlanner software has been developed which helps to generate and simulate various scenarios of radiation hazardous operations and select the best option in order to provide effective protection of the workers. Public&Environmental RP. Our experience shows that environmental remediation of the nuclear legacy sites requires the improvement of the radiation-hygienic monitoring methodology, based on the comprehensive assessment of the contamination both by radiation and by chemical factors. Databases have been developed for each nuclear legacy site. Radioecological maps were developed for the purposes of visualization of the changing radioecological situation. The results of the radiation monitoring as of 2005-2013 characterize the radiation situation in the supervision area of the NLS as normal (annual public doses are less than 1mSv), with no trend towards deterioration[2,3]. The emergency response and preparedness. In order to work out the elements of the response to radiological accidents and emergencies, emergency training was performed in 2006 at the Andreeva Bay STS and in 2009 at the SevRAO facility “Ostrovnoy”. 4. CONCLUSIONS  The development of methodology of comprehensive radiation and chemical monitoring and enhancing models to assess radiation and chemical risks are necessary. It is also necessary to consider the feasibility of establishing reference levels in terms of chemical factors  There is an urgent need to develop practical guidelines on procedure of reference levels establishment for the purpose of optimization of the public RP under various scenarios of legacy site remediation.  Рublic awareness and development of effective communication technologies to improve cooperation between local authorities and the population REFERENCES [1] SNEVE, M.K, · KISELEV, M.F, SHANDALA, N. K. Radio-ecological characterization and radiological assessment in support of regulatory supervision of legacy sites in northwest RussiaJournal of Environmental Radioactivity Volume 131, May 2014, Pages 110–118 [2] SNEVE, M.K, · SHANDALA, N. K, KISELEV, S.M, · SIMAKOV, A.V, · TITOV, A.V,· SEREGIN, V.A, · KRYUCHKOV, V.S, · SHCHEBLANOV, V.U, · BOGDANOVA, L.S, · GRACHEV, M.I, SMITH, G.M. ·Radiation safety during remediation of the SevRAO facilities: 10 years of regulatory experience Journal of Radiological Protection 07/2015; 35(3):571-596. [3] SHANDALA, N. K, KISELEV, S.M, LUCYANEC A.I, TITOV, A. V, SEREGIN, V. A, ISAEV, D. V, AKHROMEEV, S. V. Independent regulatory examination of radiation situation in the areas of spent nuclear fuel and radioactive wastes storage in the Russian far east. Radiation Protection Dosimetry 04/2011; 146(1-3):129-32.

Speaker: Dr Nataliya Shandala (SRC FMBC)

09:00 Role of Regulatory Body in Decommissioning of Research Reactors and Radiological facilities of BARC and Environmental Remediation

Abstract: Decommissioning of nuclear and radiological facilities and environmental remediation has always been a challenging task. Nuclear industry in India has completed more than six successful decades of functioning and resulted in completing useful life of few nuclear reactors and radiological facilities thus warranting serious efforts for evolving and implementing the cost effective and sustainable techniques for decommissioning and environmental remediation. Though, India has decommissioned successfully few experimental nuclear reactors and radiological facilities in recent past, still decommissioning is an area which requires international thrust and effort to standardize the methods and techniques of decommissioning and environmental remediation. In the new facilities, provisions for decommissioning are well thought and incorporated from design and inception stage itself but in older facilities these provisions were not thought of during initial days thus making decommissioning and environmental remediation task more difficult and challenging. Due to health hazards associated with radioactive wastes and activation products and their long lives, control and review by Regulatory authority becomes very important to protect the environment as per existing international norms of radiological protection. 1.INTRODUCTION: BARC Safety council (BSC) is the regulatory body for safety review of nuclear and radiological facilities involved in the field of nuclear research in Bhabha Atomic Research Centre (BARC) in India. In the ambit of BSC there are many research reactors and radiological facilities which are being designed, erected, commissioned, operated, maintained and decommissioned after completion of their useful life. 2.METHODS: A 3-tier safety framework is actively involved in pursuing the safety & regulatory review of the activities of these facilities of BARC. BSC is the apex body authorized for issuing directives. In second tier, there are various expert committees depending on their roles and responsibilities but for the activities related to decommissioning and environmental remediation, the two committees responsible are ‘Operating Plants Safety Review Committee (OPSRC)’ and Committee to Review Applications for Authorization of Safe Disposal of Radioactive Waste (CRAASDRW). These committees are assisted by Unit Level Safety Committees (ULSCs) which are seven in number depending on their domains of functioning and responsibilities. Each committee consists of 8-10 experts of their field having one Chairman and one Member-Secretary. Metallurgical Operations Safety Committee (MOSC) is the committee responsible for safety review of front end activities of nuclear fuel cycle. Safety Committee on Radiological Operations (SCRO) is the committee responsible for safety review of activities related to fabrication of advanced nuclear fuel and post irradiation examinations of nuclear fuel. Research Reactor Safety Committee (RRSC) is the committee responsible for safety review of operations and maintenance related activities of research reactors of BARC. Unit Level safety Committee-Nuclear Recycle Group (ULSC-NRG), Unit Level safety Committee- Nuclear Recycle Board- Kalpakkam [ULSC-NRB (K)] and Unit Level safety Committee - Nuclear Recycle Board- (Tarapur) [ULSC-NRB (T)] are three committees responsible for safety review of back end activities of nuclear fuel cycle for the BARC facilities at Trombay, Kalpakkam and Tarapur sites respectively. Unit Level safety Committee - Particle Accelerator (ULSC-PA) is the committee responsible for safety review of operations and maintenance related activities of the particle accelerators of BARC. In addition to the 3-tier review by regulatory authority, there is a Plant Level Safety Committee (PLSC) in every Group of BARC which assist each ULSC and review first, any issue occurred in the facility. 3.RESULTS: Decommissioning and environmental remediation are the planned activities which are implemented in field after thorough review by the group of experts in 3 tiers of safety and regulatory framework of BARC. Any proposal related to decommissioning and environmental remediation is first put up for initial safety review by the concerned facility to the respective PLSC. Proposal is taken up for review at the earliest by PLSC and along with the recommendations of PLSC; the proposal is put up to ULSC for review by the facility. Depending on the urgency, ULSC meets and deliberates on the issue and gives its recommendations. Along with the recommendations of ULSC, Facility applies to OPSRC/CRRASDRW for the higher level review. OPSRC/CRRASDRW reviews the proposal and gives recommendations. Along with the recommendations of OPSRC/CRRASDRW, Facility applies to BSC for the necessary clearances. BSC, the apex body, reviews the request and grants the necessary clearances to the facility subject to certain stipulations. Facility implements the plan and submits the status report to the regulatory body from time to time. 4.CONCLUSIONS: The role of regulatory body is of paramount importance in the activities related to decommissioning and environmental remediation in view of the fact that sometimes the jobs are outsourced and so strict quality control during implementation of plan becomes responsibility of not only facility but of regulatory body also. To achieve this, institution of periodic regulatory inspections or deputation of in-house representative of regulatory body becomes necessary.

Speaker: Mr RAJDEEP RAJDEEP (Bhabha Atomic Research centre, Mumbai, India)

09:00 Russian federal regulations and rules in the area of decommissioning: the current state and perspective for development

The existing Russian safety requirements in the area of decommissioning began to form after release of Federal Law №170-FZ "On the Use of Atomic Energy" [1] in 1995, which defined facilities decommissioning as activity in the field of nuclear energy use. Federal Regulations and Rules (FRR), which regulate safety of decommissioning, have been developed after release of Federal Law №170-FZ involving Russian nuclear regulatory body (Rostechnadzor) and its technical and scientific support organization (SEC NRS). They include: – Rules on Safety Ensuring During Industry Reactors Decommissioning (NP-007-98); – Safety rules for decommissioning of nuclear power plant unit (NP-012-99); – Safety Rules for Nuclear Research Installations Decommissioning (NP-028-01); – Rules on Safety Ensuring During Decommissioning of Ships and other Vessels with Nuclear Installations and Radiation Sources (NP-037-02); – Safety Rules for Decommissioning of NFC Nuclear Installations Decommissioning (NP-057-04); – General Safety Provisions for Radiation Sources (NP-038-11), which contain a separate section "Safety of decommissioning radiation sources". Common feature of the listed FRRs is accounting of decommissioning on the design stage of facility and acquisition of information important for decommissioning at the operational stage. Mentioned above FRRs establish requirements for: – development of decommissioning program until facility permanent shutdown; – conducting of comprehensive engineering and radiation survey; – development of facility decommissioning project; – preparation of safety report to be submitted to the Rostechnadzor to obtain the decommissioning license. Various decommissioning projects have started in Russia in last years. They include decommissioning of NPP unit and research facilities. Besides that, Russian legislation in the field on nuclear energy use was changed significantly. In particular, changes have been introduced into Federal Law № 170-FZ [1] and Federal Law № 190-FZ "On the Management of Radioactive Waste and Amendment of Some Acts of the Law of the Russian Federation" [2] was put into force. The changes have shown the need to update the regulatory framework in force in the area of decommissioning. Federal regulations and rules NP-091-14 "Safety Ensuring in Decommissioning of Nuclear Facilities. General Provisions.", which came into force in 2014, were developed on the first stage of regulatory framework updating. NP-091-14 covers all types of nuclear facilities and combines requirements common to decommissioning thereof. The IAEA documents were taken into account (in particular "Decommissioning of Facilities" GSR part 6 [3]) during development of NP-091-14. This document enshrined that decommissioning planning must be carried out through all stages of facility lifecycle, preceding it’s decommissioning. Currently the next stage of regulatory framework revision is carried out – the improvement of FRRs, listed above. FRR, regulating safety during decommissioning of radioactive waste storage facilities is under preparation. One of the requirement stated in all FRR revised, is that the operating organization must define possible decommissioning strategies at initial stage of facility’s lifecycle and select preferable strategy as result of comparing possible strategies. Implementation of final survey of facilities site after decommissioning completion to confirm that facility end-state have been met is another requirement. The immediate dismantling, the deferred dismantling and entombment (or theirs combination) are provided as possible decommissioning strategies. Entombment is allowed only for the limited number of facilities which contain radioactive waste referred by the Resolution of the Government of the Russian Federation to the "special" waste. References 1. Federal Law №170-FZ, dated 21.11.1995, "On the Use of Atomic Energy" 2. Federal Law №190-FZ, dated 11.07.2011,"On the Management of Radioactive Waste and Amendment of Some Acts of the Law of the Russian Federation" 3. INTERNATIONAL ATOMIC ENERGY AGENCY, Decommissioning of facilities. IAEA Safety Standards Series, № GSR part 6, Vienna (2014)

Speaker: Dr Anatoliy Schadilov (SEC NRS)

09:00 TURKISH REGULATION ON DECOMMISSIONING FOR NUCLEAR FACILITIES

Turkey has currently no nuclear power plant (NPP) in operation but there are two NPP projects are being implemented: Akkuyu NPP and Sinop NPP. The purpose of the draft regulation on decommissioning for nuclear facilities is to regulate the rules to be applied during safely decommissioning of nuclear facilities. This draft Regulation covers activities and roles in the decommissioning process of nuclear power plants. Responsibilities of the operating organizations are also regulated in the intergovernmental agreements including the financial issues which are also given in this paper.

Speaker: Mr KEMAL DOĞAN (Turkish Atomic Energy Authority)

Paper IAEA-CN-238-13P_TURKISH_REGULATION_ON_DECOMMISSIONING_FOR_NUCLEAR_FACILITIES.docx

11:00 → 11:30 Coffee break

11:30 → 13:30 Session 2 - 2: Regulatory Framework and Standards for D&ER 2

The purpose of this session is to discuss significance of regulatory requirements to implementation of D&ER programmes, including discussion of evolution of regulations and regulatory approaches.

11:30 Regulatory Requirements for Decommissioning of Nuclear Facilities in Germany

In Germany the regulatory framework is designed to be applicable to all types of nuclear facilities / activities including the decommissioning of nuclear facilities. This generic approach has been developed for regulating nuclear activities over the many years of nuclear experience. Provisions applicable to an individual subject area, such as decommissioning of nuclear facilities, are “scattered” in the various documents of the national regulatory framework. In order to provide assistance to those who are dealing with decommissioning of nuclear facilities, the Federal Ministry for the Environment, Nature Conservation, Building and Nuclear Safety (BMUB) issued a “Decommissioning Guide”, which compiles all regulations relevant for licensing and supervision of nuclear facilities in the decommissioning phase. In addition, the Nuclear Waste Management Commission (ESK) has published guidelines, which are complementary to the Decommissioning Guide in a technical sense. Both documents have been updated in the light of the challenges resulting from the increasing number of nuclear power plant decommissioning projects in Germany. The paper provides an overview on the regulatory framework for decommissioning and highlights the recent updates of relevant guiding documents.

Speaker: Dr Bernd Rehs (Federal Office for Radiation Protection)

11:45 Remediation of legacy low-level radioactive waste: Regulatory oversight for two major remediation projects in Canada

This paper provides a summary of the Canadian Nuclear Safety Commission’s (CNSC) regulatory oversight process for licensing two environmental remediation projects of legacy radioactive waste in Canada. The CNSC, which reports to Parliament through the Minister of Natural Resources, is the sole regulatory authority for the peaceful use of nuclear energy in Canada. As Canada's nuclear regulator, the CNSC is responsible for licensing of all aspects of uranium mining, including remediation activities at legacy sites. Staff of the CNSC has a great deal of experience in reviewing documentation in support of a licence application. Licence applications are reviewed by CNSC staff, whose role, in general, is to provide authoritative and defensible advice and recommendations to the Commission. Although the licensing process for legacy sites is no different than for any other CNSC license, assuring regulatory compliance requires the addition of unique elements such as site characterization of an existing site, clean-up criteria, community concerns, historic records and long-term management. These are some of the challenges presented in the remediation of historic waste, and hence the regulation of such sites. The purpose of the regulatory review of a licensing document is to confirm that sufficient information is provided to justify claims that regulatory requirements have been or will be met. In doing so, the regulatory review may need to corroborate claims of safety, performance or environmental protection made in the submission. The advice and recommendations of CNSC staff to the Commission regarding a licensing decision arise from staff’s critical review of applications and licensees’ submissions, activities and actions. After a licence has been issued for the remediation of a contaminated site, there is often a need to review additional documentation that validate compliance with a regulation or licence conditions. This paper presents CNSC staff’s current regulatory review process using two examples of active Canadian nuclear legacy site remediation projects. The first project, known as the Port Hope Area Initiative or PHAI (www.phai.ca) involves the federal clean-up of historic low-level radioactive waste situated in the municipalities of Port Hope and Port Granby in Ontario, Canada. The Government of Canada has invested over $1.28 billion (Cdn. dollars) to construct two new long term waste management facilities designed to accommodate approximately 1.65 million cubic meters of legacy contaminated wastes. PHAI is licensed under two 10-year CNSC Waste Nuclear Substance Licences issued by the Commission in 2011 and 2012. The second project is the remediation of the Gunnar mine and mill site located in northern Saskatchewan. The Gunnar mine and mill was operated by the former Gunnar Mining Limited from 1955 to 1963 and was decommissioned in 1964. At decommissioning, the open pit and underground workings were flooded, and the mine shaft and openings were plugged with concrete. The buildings at the site were demolished in 2010. The tailings, waste rock, and other mine waste were left behind and remain on the surface to this day. This remediation project is also licensed under a 10-year CNSC Waste Nuclear Substance Licence. Large remediation projects such as those discussed in this paper require the conduct of an environmental assessment prior to a licensing decision. The proponent is then required to submit licensing documentation to support a licensing review process under the Nuclear Safety and Control Act (NSCA) that, if successful, would lead to the issuance of a licence to authorize the remediation work. These legal requirements rely on the submission of documentation that is reviewed by CNSC staff and presented to the Commission for consideration. If the remediation project is approved, then verification activities are conducted by CNSC staff to ensure that the project proceeds as planned and that all measures related to safety and to public engagement are appropriately implemented. Key aspects in CNSC staff’s current regulatory oversight are highlighted such as the assessment of data, consideration of limitations and supplications, long-term monitoring, community concerns and environmental objectives for contaminated sites. This document shares the experiences of the Canadian nuclear regulator’s regulatory review process to evaluate remediation projects and is intended to encourage discussion about the regulation of remediation projects and to help those with less experience in licensing remediation projects.

Speakers: Mr john thelen (canadian nuclear safety commission), Ms karina lange (CNSC)

13:30 → 15:00 Lunch break

15:00 → 16:00 Session 3 - 1: Decision-Making Process: Social and Stakeholder Involvement during the Lifecycle of Programmes 1

The purpose of this session is:
• To demonstrate the importance of stakeholder engagement in obtaining the necessary social licence for D&ER activities
• To highlight how experience with stakeholder involvement over the past 10 years has impacted decision-making approaches for D&ER

15:05 SOCIETAL CONSTRAINTS RELATED TO STAKEHOLDER INVOLVEMENT IN ENVIRONMENTAL REMEDIATION AND DECOMMISSIONING PROCESSES

Different academic studies and the IAEA project CIDER (Constraints in Decommissioning and Environmental Remediation) argue that lay people are able to reason about complex technical matters. Judgement of radiological risks related to contaminated environment or decommissioning of nuclear installations includes a wider range of considerations in lay population reasoning processes as well as in experts. Judgment includes, not only scientific and factual knowledge about ionizing radiation, but also values, trust, experiences, familiarity with risk, etc. Therefore, risk communication about decommissioning and environmental remediation processes (D&ER) should not be seen as a form of technical communication and education whereby the public should be informed about remediation and decommissioning plans nor as a marketing practice with the aim to persuade people to adopt certain solutions. Nowadays, risk communication is seen as a stakeholder engagement process through which the gaps between stakeholders can be bridged. Different past experiences of IAEA Member States point out diverse constraints related to stakeholder involvement in D&ER processes, leading to unsuccessful remediation of the environment or decommissioning. The purpose of this presentation is to highlight the major societal constraints that some organisations in different IAEA Member States may encounter when implementing D&ER programmes and different approaches to overcome these constraints. The following constraints are discussed: i) limited technical knowledge and understanding of the problem and process by stakeholders; ii) groups and individuals opposing project implementation, iii) the ‘Not In My Backyard’ (NIMBY) syndrome - with particular importance for the disposal of generated wastes from D&ER operations, iv) different demands and concerns between stakeholders, v) limited budget to cover stakeholders demands, vi) negative experience with previous D&ER programmes, vii) lack of support by the Governmental authorities to implement D&ER, viii) changing administrative procedures and legal framework, ix) lack of trust between stakeholders, x) lack of recognition of links between environmental, economic, and social concerns. Examples of existing practices related to the integration of societal aspects in D&ER programmes worldwide collected in the context of the IAEA project CIDER are presented in the light of recent trans-disciplinary research findings. Additionally, a CIDER initiative on a stakeholder engagement programme taking place in real uranium project in Brazil will be presented stressing lessons learnt which could be generally used.

Speaker: Dr MERITXELL MARTELL (MERIENCE)

15:35 ABORIGINAL AND NORTHERN INVOLVEMENT AND BENEFITS FROM GUNNAR URANIUM MINE ENVIRONMENTAL REMEDIATION – NORTHERN SASKATCHEWAN

ABORIGINAL AND NORTHERN INVOLVEMENT AND BENEFITS FROM GUNNAR URANIUM MINE ENVIRONMENTAL REMEDIATION – NORTHERN SASKATCHEWAN MULDOON, Joseph; CALETTE, Mark; and SCHRAMM, Laurier L., Saskatchewan Research Council, Saskatoon, SK, Canada E-mail address: muldoon@src.sk.ca 1. INTRODUCTION – MINE SITE AND HISTORY The Saskatchewan Research Council (SRC) was contracted by the Province of Saskatchewan to manage the remediation of 37 abandoned uranium mine sites in northern Saskatchewan with the Gunnar mine and milling site being the largest and most complex. Decommissioning was limited to flooding the pit and capping the mine shaft. Risk assessments were carried out during the licence application and review processes and in 2010 it was determined that the degraded nature of the buildings on-site posed an unacceptable physical hazard to public safety. These buildings were demolished in 2011/12 with debris stored on the original footprint for final disposal once the environmental assessment and licensing processes were completed. Due to the limited information available on this legacy mine site, it was not economically feasible to collect all required technical information to fully inform the environmental review and licensing processes. As a result, in close consultation with the regulatory agencies, SRC developed a decision tree approach that allowed a sequenced determination of detailed remediation plans for the four major components of the site: tailings areas, waste rock, demolition debris and associated materials, and the open pit. The environmental assessment and licensing approvals were obtained in 2014/15. 2. SRC’S APPROACH TO ENGAGEMENT/INVOLVEMENT The most important component of engagement is that of commitment from the entire organization both top-down and bottom-up. As an example, before the contract was signed between the Province of Saskatchewan and SRC, the President and CEO of SRC personally became involved with the project meeting with the key community members and in the meetings that established a Project Review Committee comprised of designated community representatives. This committee was established well in advance of any regulated timeframes for community involvement. A full-time position was established within the SRC remediation team to collaborate with both SRC employees and the northern communities. SRC employees have become fully invested in this process which has resulted in a bottom-up program development and continuous improvement. 3. INITIAL COMMUNITY INVOLVEMENT ACTIVITIES Multiple and meaningful interactions starting at the initial planning stages and continuing through the setting of project objectives, project review and approval stages, implementation and post-monitoring are crucial to successful community engagement. To the end of 2015, SRC has held over 130 community meetings with northern people living in the region. It is critical in these processes to ensure that input from these interactions is obtained and actions taken within the project development and implementation that demonstrates active listening. 4. COMMUNITY ECONOMIC INVOLVEMENT AND DEVELOPMENT The direct involvement of communities in the delivery of services required to remediate the Gunnar mine site was/is a critical component of successful engagement. SRC’s procurement team developed detailed criteria for selecting the contractors that would deliver the various components of the remediation project. The criteria included significant weighting to ensure local residents would be trained, employed in, and benefit from the remediation work. To assist with this process a designate from the area was appointed to the procurement decision-making panel. Advance training was provided to not only assist the local residents gain employment in specific remediation work but to equip them to be successful pursuing other similar employment opportunities once the remediation work was/is complete. Community liaisons were designated (through the contractors) to assist in the development and designation of employment opportunities. Some Northern people are being trained in monitoring to ensure longer term project involvement. A mentorship program has also been developed to help train post-secondary Aboriginal students in science and engineering thereby increasing the number of highly qualified people that could be engaged in the remediation work. 5. CURRENT STATUS OF PROJECT The phase II license, which allows the remediation to begin, is expected in late 2015/early 2016. The remediation plans for the other site components will be finalized and implemented in a coordinated fashion to maximize community engagement as well as program efficiency and effectiveness. 6. SUMMARY In order to be successful in meaningful community engagement, a total commitment to the process must be developed and maintained. Capacity, patience and genuine goodwill determine success. Input must be collected and actioned in ways that are evident to those who participate and those directly impacted by the project activities. Employees involved with the project must feel engaged and empowered to carry out the engagement activities and to be fully successful must be motivated and passionate. In all stages of the project implementation and post-monitoring, community engagement and direct involvement in the project will remain a high priority.

Speaker: Dr Joseph Muldoon (Saskatchewan Research Council)

15:00 → 18:00 Session 3 - Poster

15:00 REMEDIATION STRATEGIES AND DECISION-AIDING TECHNIQUES

Considerable experience has been gained in remediating areas contaminated by nuclear past activities and accidents. Despite the technical and scientific experience gained from this, environmental remediation remains a lengthy and expensive process, emphasizing the need for proper justification and optimization of the preferred remediation approach. The contamination can include areas with elevated levels of naturally occurring substances, or former industrial sites which may have left a legacy of contamination from operational activities, as might be found in sites being decommissioned. It can also include land directly or indirectly contaminated as a result of accidents, spillages, aerial deposition and migration. The need for remediation is defined basically by two questions: is there a risk due to the contamination, and if there is a risk then how can the contamination be dealt with in a safe manner, whilst taking along those interested parties who are either impacted directly by the hazard or those with a legitimate interest in the outcome? The focus of this paper is on the decision making process associated with the assessment of remediation options. The steps prior to a decision being made are critical to the success of a remediation project. Several tools – stand-alone or linked together – are available for risk and socio-economic analysis that help in defining the remediation approach. Effective and supported decision making requires a process that is transparent and straightforward, and many techniques have been developed to facilitate stakeholder discussion, encode uncertainties, elicit value judgements and which guide decision makers and interested parties to a shared understanding of the costs, risks and benefits of different potential remediation options and strategies. Each remediation project is going to have its own set of unique factors that need to be considered. These processes can use different ‘societal’ decision analysing tools, bringing together Science (our knowledge and understanding of the context and the possible consequences of our actions, including that of doing nothing) and Values (costs, benefits, ethics, anything that bears on how much particular consequences matter to us). Depending on the complexity of the remediation project a range of decision aiding tools can be used. Formal decision analysis, including multi-criteria decision analysis (MCDA), is an approach that has assisted decision making in complex contamination problems. Tools based on MCDA, and other methods, offer a structured approach to determining the preferred remediation option. A generic framework will be presented on how to plan a remediation project, to gather the appropriate technical and non-technical information, how interested parties should be managed in the assessment of the remediation options and to identify the tools to facilitate a sound decision. It is important to recognize that there is no single method or approach that can be used across all problems and all decision models are only tools for exploring decisions: they don’t make the decision! When a decision to remediate a contaminated area is required, then a remediation plan should be prepared by the remediation project manager. This plan should outline the steps to be taken for gathering appropriate technical information and engaging interested parties. Decisions are likely to be more widely accepted if it the process is open and transparent and considers both technical and non-technical factors in an appropriate and meaningful way., and the paper seeks to offer advice to remediation project managers on the factors they should consider when seeking to make a decision. This work has been achieved within the MODARIA (Modelling and Data for Radiological Impact Assessments) Programme of the AIEA from 2012 to 2015. The general aim of the MODARIA Programme is to improve capabilities in the field of environmental radiation dose assessment by means of acquisition of improved data for model testing, model testing and comparison, reaching consensus on modelling philosophies, approaches and parameter values, development of improved methods and exchange of information. Within this context, it was deemed appropriate to explore how decisions are made, which often use the output from modelling. This was undertaken by the Working Group 1 of MODARIA, dedicated to “Remediation strategies and decision aiding techniques”.

Speaker: Mr Thomas LE DRUILLENNEC (EDF)

15:00 STAKEHOLDER INVOLVEMENT IN NUCLEAR POWER PLANT (NPP) AND RESEARCH REACTOR (RR) DECOMMISSIONING

Abstract: Decommissioning of a nuclear power plant (NPP) or a research reactor (RR) in Germany requires a decommissioning license. As part of the licensing procedure a stakeholder involvement is mandatory by applicable laws. The experiences from past and present stakeholder involvement processes during the licensing procedures are presented and the influence of stakeholders on decommissioning projects in licensing and the pre-licensing procedures is discussed. 1. INTRODUCTION To decommission a NPP or a research reactor (RR) it is necessary to apply for a decommissioning license. As part of this licensing procedure a stakeholder involvement is mandatory by applicable laws. There are some examples where either the facility owner or the regulatory body started a stakeholder involvement initiative before a decommissioning license was applied for. 2. LEGAL BACKGROUND The Atomic Energy Act (AtG) [1] states, that to decommission a licensed facility where fission material was generated, handled or fissioned a license for decommissioning is needed. The Nuclear Licensing Procedure Ordinance (AtVfV) [2] regulates the licensing procedure, including the involvement of stakeholders, and the Environmental Impacts Assessment Act (UVPG) [3] determines the implementing of an environmental impact assessment. The AtVfV also rules the announcement and display of documents, the raise of objections and the holding of a hearing. This hearing is an oral, non-public review of objections where stakeholders can explain and state more precisely their timely submitted objections to the regulatory body. After this review the objections are reevaluated and either rejected or have to be considered during the licensing procedure. 3. EXAMPLES In the stakeholder involvement of the licensing procedure for decommissioning of NPP Stade in 2003 only five different objections were submitted. All were rejected by the regulatory body [4]. During the stakeholder involvement of the licensing procedure for decommissioning of NPP Obrigheim in 2008 897 persons objected the construction of the interim storage for spent fuel on-site, but there were no objections against the decommissioning itself [5]. In 2012 a voluntary stakeholder involvement in the licensing procedure for the 3rd decommissioning license of NPP Obrigheim was performed [6]. There were 72 (combined) objections against this license and a huge public discussion. After final shutdown of NPPs Philippsburg 1 and Neckarwestheim I in 2012 two information commissions were established under the initiative of the state government [7] to inform the local stakeholders and address their concerns. Since then there were a number of gatherings of each commission where different aspects were brought to the agenda (e.g. decommissioning strategy, radioactive waste treatment …). After final shutdown of the RR Geesthacht in 2010 the operator initiated in 2012 a dialogue with local stakeholders, parties and NGOs [8]. For this dialogue a renowned mediator was contracted. A small group of different stakeholders and the operator agree on the agenda for the meetings. Different topics were covered until now (e.g. licensing procedure, decommissioning aspects, waste management …). Public involvement for the decommissioning license of NPP Unterweser and for the handling license for construction and operation of an interim storage for radioactive waste is ongoing [9]. Due to the necessary application for a building permit for the construction of the interim storage it was deemed necessary to broaden the stakeholder involvement and publicize the applications for decommissioning license and for the construction in more places and online according to building legislation. 4. CONCLUSIONS Stakeholder involvement as required by Atomic Energy Act is restricted to objections of stakeholders to the application for decommissioning of nuclear installations. The regulatory body has to verify these objections and to take them into consideration in the decommissioning license. Examples of voluntary stakeholder involvement processes exist, which were either initiated by the local administration or the facility owner.

Speaker: Dr Gerd Bruhn (Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH)

15:00 URANIUM MINING LEGACY SITES AND ENVIRONMENTAL REMEDIATION IN PORTUGAL: REVIEW OF PROGRESS ACHIEVED

Abstract: A vast legacy of radium and uranium production in Portugal is described. By 2005 the conclusions of a radiation risk assessment carried out at uranium legacy sites concluded that enhanced exposure existed in some areas and recommended environmental remediation action. A remediation program was approved, funded and it is implemented since. An overview of results achieved in 10 years is presented.

Speaker: Prof. Fernando P. Carvalho (Instituto Superior Técnico/Laboratório de Protecção e Segurança Radiológica,)

16:00 → 16:30 Coffee break

16:30 → 18:00 Session 3 - 2: Decision-Making Process: Social and Stakeholder Involvement during the Lifecycle of Programmes 2

The purpose of this session is:
• To demonstrate the importance of stakeholder engagement in obtaining the necessary social licence for D&ER activities
• To highlight how experience with stakeholder involvement over the past 10 years has impacted decision-making approaches for D&ER

16:30 Experience of Environmental Remediation in Date City

On March 11, 2011, I got the news on TV that the Fukushima Daiichi NPS had got into a critical situation after the massive earthquake and tsunami. At that time, I understood that there would be no impact on Date City, 60km away from the station, and informed that to citizens. However, the SPEEDI results that were suddenly made public for the first time 10 days after the accident revealed that Date City would also have the possibility of radioactive contamination. We did not have any practical knowledge or measurement devices of radiation. We did not get any concrete instructions either from Fukushima Prefecture or from the national government in that unprecedented emergency situation. Thus, I decided to take measures on my own judgment and considered evacuation and decontamination, reviewing the records of the Chernobyl accident. By the mid-April of 2011, public anxiety about health effects of radiation on children had grown. However, the government instruction at that time was simply to keep children indoors if the radiation dose rates exceeded 3.8μSv/h (20mSv/a) on the school grounds. We consulted experts, and based on their suggestion, took our own emergency actions including decontamination of all the school grounds and swimming pools, and provision of glass badges to all students of the elementary and junior high schools in the city. We took all these emergency measures with our own decision and allocated budget of JP\1.0B. An evacuation order was issued to our neighboring village. As it turned out that the adjacent area in Date City also had high radiation level, we helped the residents to evacuate in June 2011. However, evacuation is an urgent measure. In view of considerable detriments of evacuation, I decided that we should decontaminate the living area immediately. For immediate decontamination, we divided the city into three zones (A, B, and C) based on the monitoring results and implemented decontamination according to the level of radiation. That decision led to the completion of decontamination of about 23,000 houses in about two years and minimized the cost, about JP\25B(JP\15B was spent for decontamination of 2,600 houses in the zone A with the highest radiation level among the zones). The largest challenge in decontamination was how to secure temporary storage sites for generated soil and wastes, due to the NIMBY problem. We held a series of discussions with residents in each community. Ultimately, the city has had more than 100temporary storage sites. In parallel with decontamination, we provided glass badges to 15,000children since June 2011 and to all the 65,000citizens since July 2012to manage their exposure and to mitigate their concerns. One-year accumulated data till June 2013 were collected from about 50,000 people. The data show that annual additional exposure was less than 1mSv for about 66%and less than 2mSv for about 94% of them. While our practical criterion for remediation was 5mSv/a, nobody below 15 years old received more than 4mSv/a. Thus, our actions for children can be accounted a success. We also provided all citizens with WBC examination and food inspection whenever needed. As of July 2015, the average additional exposure dose of citizens is 0.59mSv/a, and 84.3%of citizens receive less than 1mSv/a additionally. The situation can be regarded as safe; however, concerns still remain among some people. Their long-lasting concerns can be attributed to the national policies including the decontamination target and food regulation. The Government set the long term decontamination goal as the additional dose of 1mSv/a in November 2011, still in the midst of the accident response. It also set the standard limits of radionuclide in food as 100Bq/kg in April 2012, by revising the former tentative criterion 500Bq/kg. These hard targets may have further increased people’s anxiety. It is also an issue that the long term target of 1mSv/a tends to be mistaken as if it has to be immediately achieved to secure safety. The Government also converted the annual additional exposure of 1 mSv into the air dose rate of 0.23μSv/h. 0.23μSv/h is a criterion to designate survey areas but has been mistaken as an absolute standard for decontamination. Our data show, however, that the annual additional exposure doses fall below 1mSv even if the air dose rates are around 0.50μSv/h. Based on the fact, I made a proposal that the Government should review the criterion of 0.23μSv/h so that municipalities focus on truly required decontamination. However, the Government has been reluctant to formally accept it. We have made certain progresses in radiation protection; however, we still face challenges to address people’s persistent anxieties and reputational damages.

Speaker: Mr Shoji Nishida (Mayor of Date City)

Wednesday, 25 May

09:00 → 11:00 Session 4A - 1: Technical and Technological Aspects of Implementing Decommissioning Programmes - Parallel Session

The purpose of this session is:
• To review progress in decommissioning technologies and cost estimation over the past decade and to identify challenges and needs in the future
• To review current developments, including from damaged facilities, with a view to identify existing gaps in knowledge and needed improvements

09:10 STRATEGIES FOR OPTIMISATION OF THE RADIOLOGICAL CHARACTERISATION OF A NUCLEAR FACILITY

Radiological characterisation plays an important role in the process of decommissioning shut-down nuclear facilities in order to ensure protection of the environment and radiation safety. It is a key element for planning, controlling and optimising the decommissioning and dismantling including the residual materials and waste management. At all stages of a decommissioning programme or project, adequate radiological characterisation is crucial. Experience has shown that data and information from the operational phase of a nuclear facility can - beside data and information collected and analysed for related decommissioning activities - be fundamental for decisions on waste management and for characterisation of radioactive waste. Some information may be hard, costly or even impossible to obtain at later stages in the waste management process once the dismantling has taken place. This was the reason why the Working Party on Decommissioning and Dismantling (WPDD) of the OECD Nuclear Energy Agency (NEA) has decided to establish a Task Group on Radiological Characterisation and Decommissioning (TGRCD). The Task Group has completed a first phase that was focused on overall strategies of radiological characterisation in decommissioning and is in the middle of its second phase with the focus on nuclear facility characterisation from a waste and material end-state perspectives. This paper summarises the activities performed and to be performed by the Task Group.

Speaker: Mr Arne Larsson (Studsvik, Sweden)

09:35 LESSONS LEARNED FROM PREPARING DECOMMISSIONING OPERATIONS WITH VIRTUAL REALITY

After having operated numerous nuclear facilities since the 1950s, the CEA (French Atomic and Alternative Energies Commission) must now manage the dismantling of those which have reached the end of their lifetime. These high priority actions have led to the creation of an R&D dismantling division which aims at providing innovative tools, including intervention scenario simulation. Simulation is a good means of visualizing highly radioactive environments where humans cannot enter, of testing different technical alternatives, and of training workers prior to interventions. For a few years, the CEA has developed a generic simulation platform based on virtual reality (VR) technologies, usable on any decommissioning project. On this platform, different kinds of simulation can be run: physics, kinematics, virtual human simulation and dose-rate calculation. All these modules are embedded in a software called iDROP, taking into account the whole aspects involved in nuclear operations in a single simulation. This paper describes the different application cases where VR simulation has been used to design dismantling operations, presents the lessons learnt from these different implementations.

Speaker: Mrs CAROLINE CHABAL (CEA)

10:00 DECOMMISSION OF ‘NUCLEAR LEGACY SITE AT VNIINM

Extensive work on the decommissioning of own research facilities are conducted in A.A.Bochvar Research Institute of Inorganic Materials (VNIINM). Since 1946 these facilities used for Plutonium, Uranium and other radionuclide chemistry, nuclear fuel research and reprocessing, radioactive waste and other technologies. Amount of contaminated structures and unused old equipment, including unsorted radioactive waste onsite within Moscow area required immediate decommission project to solve ‘Nuclear legacy site’ problem. The main challenges the Institute facing are the decommissioning of a number of research buildings and auxiliary buildings to the rehabilitation of the sites they host, removal from the territory of the Institute of all nuclear and toxic materials which are not subject to further use, minimization of the quantities of radioactive waste from the output of operation stages of their arising (use of low-waste technologies) and treatment (including collection, sorting, processing and conditioning) The distinctive features of the work of decommissioning radiation hazardous objects of VNIINM are a wide variety of bench equipment in research buildings, radioactive contamination of part of the rooms and equipment mainly with alpha-emitting radionuclides, significant depreciation of engineering systems located in buildings (ventilation, special drainage system and others), the presence in the research buildings of accumulated different water and organic radioactive solutions and chemical reagents requiring conversion in RW category and conditioning before transport to specialized organizations, location of VNIINM inside residential area in Moscow. The primary challenges facing the institute are decommissioning of building B and the semi industrial plant U-5. Building B was used as experimental basis for USSR radiochemical industry since 1946. Work on the decommissioning building B was started in 2008. Work is scheduled for completion in 2015, eliminating the building B. Main stage of the project covering removal of chemical reagents, decontamination, dismantling and removal of equipment, engineering systems, waste management and decontamination units is finished. Now works are on Final stage: building is dismantled, site rehabilitation activities are in progress to be finished at the end of 2015. Measures for reducing the environmental impact during the decommissioning works included dust suppression by creation of aerosol screens during dismantling, use of polymeric coatings for fixing radionuclides on radioactive contaminated surfaces, creation of “cocoon” made from a special coating on the line of scaffolds on the stage of dismantling of the building , zoning of “dirty” and “clean” areas for the prevention of secondary radioactive contamination, maintenance of supporting engineering systems for the whole period of works prior to the dismantling of the building, creation of a centralized automatized radiation control system of the institute, minimization of secondary radioactive waste. Report covers all stages of decommission project of Building B, including characterization, planning, implementing, waste management and overall performance with lessons learned. Work on the decommissioning of the plant U-5 was started in 2010. Work is scheduled for completion in 2018, eliminating the plant. Currently, performing preparatory stage works involving sanitary unit construction, radioactive and clean waste units construction, loading area construction as well as licensing decommission activities. Report covers practical feedback from performed events, including R&D and use of special technologies - decontamination polymer coatings during dismantling. Performing at VNIINM decommissioning work is associated with the solution of complex engineering and technological problems, due to specific objects and their location in dense urban area in Moscow. The experience gained here can be used in similar works on other objects.

Speaker: Mr Sergey Savin (VNIINM \ Rosatom)

10:25 ROBOT CHALLENGES FOR NUCLEAR DECOMMISSIONING OF FUKUSHIMA DAIICHI NUCLEAR POWER STATION

We will present robot technologies for nuclear decommissioning developed by IRID (International Research Institute for Nuclear Decommissioning, JAPAN). In the lead-up to retrieving fuel debris from Units 1-3 at Fukushima Daiichi, various tasks are planned to take place inside the reactor buildings. In order to perform these tasks smoothly, improving the working environment is essential. An overall reduction in radiation levels is sought through a combination of decontamination, shielding, and removal of radioactive sources. As one of key challenges, we developed technologies for remotely operated decontamination inside reactor building. Three types (high pressure water jet, dry ice blast, and suction/blast) of decontamination technology were selected. Three types of decontamination systems for lower parts of the first floor, for higher parts of the first floor, and for upper floor of the reactor building (2nd and 3rd floors) were developed respectively. A mockup test facilities were constructed and verification tests took place. The performance of decontamination, traveling, and operability and safety functions were evaluated during verification tests. It is estimated that reactors cores in Units 1-3 have melted and fuel has partially fallen with reactor internals into the Reactor Pressure Vessel (RPV) and Primary Containment Vessel (PCV). In particular, it is possible that after fuel debris melted through the bottom of the PCV in Unit 1, they emerged from the inside og the pedestal (which supports the PCV) and spread out of the pedestal opening. However the actual condition has not been confirmed. Before the fuel debris removal, it is important to know the condition of inside RPV and PCV. Because of severe environment of high radiation dose and high humidity, remotely controlled systems are indispensable. As one of challenges, we developed two systems for investigation inside the PCV. The one is technology for accessing the pedestal exterior. A shape-changing robot to investigate the pedestal exterior inside the Unit 1 PCV (the grating outside of the pedestal) was developed. The other is technology for accessing inside the pedestal. A self-miniature robot will enter via the opening of the X-6 penetration (PCV penetrating part) and after traveling through a guide pipe inserted into the PCV traverse the CRD rail into the pedestal interior. Finally we will conclude our achievements and discuss the lessons learned.

Speaker: Dr Tetsuo KOTOKU (International Research Institute for Nuclear Decommissioning (IRID))

09:00 → 18:00 Session 4A - Poster

09:00 AN APPROACH OF THE INITIAL DECOMMISSIONING PLANNING FOR THE BRAZILIAN NUCLEAR RESEARCH REACTORS IEA-R1 AND IPEN/MB-01

Abstract: The Brazilian Nuclear program has progressed in the past 60 years, resulting in six reactors, being four nuclear research reactors and 2 nuclear power plants. There is one more NPP under construction. With the establishment of the Brazilian Nuclear Energy Commission CNEN (Comissão Nacional de Energia Nuclear), in 1962, it was possible to tandardize the use of nuclear energy for peaceful purposes, regulating the activities and practices that use radiation or radioactive material. Regarding nuclear facilities, CNEN has specific legislation with various norms covering the licensing of nuclear installations and all necessary safety procedures for the operation nuclear facilities. However, the decommissioning has not been addressed, as a whole, by the current legislation. To comply with the intrinsic requirements of a nuclear plant shut down procedures, specifically the power reactors, only a resolution was published on nuclear power plants decommissioning in 2012.The aim of this work is to present the current stage of the study for the development of a preliminary decommissioning plan of the nuclear research reactors IEA-R1 and IPEN/MB-01. 1. INTRODUCTION The present work shows an approach of the studies of the initial decommissioning planning of the research reactors IPEN/MB-01 and IEA-R1, both in operation at the Nuclear Energy Research Institute – IPEN in São Paulo city. -The reactor IPEN/MB-01: it is a zero power type with maximum nominal power of 100W. The main characteristic is its versatility, allowing different core critical arrangements. The reactor has a main purpose carry out experiments and neutronic simulations of moderate reactors with light water, allowing the experimental verification of nuclear parameters. -The reactor IEA-R1: it is a pool type with a maximum power rating of 5 MWth, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor is located in a multidisciplinary facility which has been consistently used for research in nuclear and neutron related sciences and engineering. The reactor has also been used for training, radioisotope production for industrial and nuclear medicine applications, and for general irradiation services. 2. METHODS The methodology used is based on national standards and international recommendations. Several aspects have been considered to define the initial decommissioning strategy. They are: the radioisotopic inventories, the general analysis of the main factors affecting the decommissioning strategy selection (human resources, required technology, waste management in plants and health, safety and environmental impacts) and the international experience and strategies in research reactors decommissioning. -The reactor IPEN/MB-01 has as a potential source of generation of radioactive waste the critical assembly due to be more susceptible to the neutron flux which has been considered the main component on this study. -The reactor IEA-R1, has been in operation since 1957. Due the long operation time, the large amount of radioactive material come from the pool walls constructed by about 270m3 of high density concrete and 350m3 of ordinary concrete. Part of this concrete are exposed to neutrons and gamma radiation, arising during the operation of the reactor and shall be treated as radioactive waste. The study attempts to make an estimative about the amount of the concrete wall will be treated as radioactive waste arising from the possible dismantling of the reactor. 3. RESULTS -Reactor IPEN/MB-01: The results reveal that the material activities are considered low and most of the present radionuclides half-life is shorter than 100days, with rapid decay. The main components to be considered are the guide tube, the fuel rod coating and the intermediate spacing plate due to be closer to the core active region. -Reactor IEA-R1: The results show that the estimated amount of radioactive waste generated in decommissioning the pool right after reactor shutdown is 143m3. After 5 years or more this value decreases to 71m3. 4. CONCLUSIONS The present work come to contribute as a reference to develop decommissioning plans for other Brazilian facilities. In the world there are several cases reported about research reactors that after its definitive shutdown had preserved their facilities and converted into museums or for other purposes which do not resulted in the dismantling of reactor building. It can be a good alternative for the IPEN/MB-01 and mainly the IEA-R1 once it was the first nuclear research reactor built in Brazil. REFERENCES [1] COMISSÃO NACIONL DE ENERGIA NUCLEAR, “Descomissionamento de Usinas Nucleoelétricas (CNEN-Resolução Nº 133)”,(2012), Brazil. [2] INTERNATIONAL ATOMIC ENERGY AGENCY “Standard Format and Content for Safety Related Decommisioning Documents, (Safety Report Series Noº 45 )” (2005),IAEA, Vienna. [3] INTERNATIONAL ATOMIS ENERGY AGENY “Decommissioning of research reactors: Evolution, State of the art, Open issues, (Technical Report Series 446)” (2006), IAEA, Vienna.

Speaker: Dr ALVARO CARNEIRO (IPEN/CNEN Brazilian Nuclear Energy Commission)

09:00 APPLICATION OF THE IAEA DECOMMISSIONING RISK MANAGEMENT GUIDELINES DURING THE UPDATE AND REVIEW OF THE INITIAL DECOMMISSIONING PLAN OF THE SAFARI-1 RESEARCH REACTOR

Risk management process is a matured process that has been widely applied in various industries. The application of this process in Decommissioning and Dismantling has been limited. In the past few years the International Atomic Energy Agency (IAEA) launched a 4 year international project on Decommissioning Risk Management (DRiMa) with the objective of providing guidelines on how to implement risk management in decommissioning. This project was completed in 2015 and the outcomes are still in draft. The purpose of this paper is to report on the application of the DRiMa methodology during the review and update of the initial decommissioning plan of the Safari-1 research reactor. The successes and challenges experienced during the application of the risk management process will be shared.

Speaker: Mr STEVEN DHLOMO (SOUTH AFRICAN NUCLEAR ENERGY CORPORATION)

09:00 APPROACH TO SAFETY ASSESSMENT FOR DECOMMISSIONING ACTIVITIES IN SLOVAKIA

The aim of safety assessment is to identify possible safety hazards related to planned activities of the NPP decommissioning, to identify possible operational events and their assumed consequences to personnel and public. Moreover it is necessary to demonstrate that in the case of the most serious assumed operational events, consequences will be mitigated by defined manner and any non-reasonable hazards to personnel or public will not take place. Paper describes approach to safety assessment for decommissioning activities in Slovakia. Presented methodologies for hazards identification and other safety analyses are applicable for planned decommissioning activities as well as for operation of waste management facilities.

Speaker: Dr Frantisek Ondra (DECOM, a.s.)

09:00 Assessment Gamma Dose Rate for Hypothetical Radwaste Container

Metallic solid radioactive waste is the main type of radioactive waste generated from decommissioning operations. Transport, storage and disposal regulations require for gamma emitting radioactive waste ( mainly by Cs-137 isotope), that the dose rate in the proximity of the container should stand below a certain threshold. Also, the conditioning technique ( using cementation technique ) based on certain matrix with specific ratios should be able to alternated the gamma radiation activity to the minimum level or to acceptable dosage rate on the contact or meter distance from the container. In this paper; to assess dose rate in safe way, assumption based on metallic pieces waste were polluted with (Cs-137) were conditioned with cement matrix contained in carbon steel drum volume 220 liter ,60cm diameter a and dose rate measurement applied in vicinity of the container. Instead of real polluted metal waste, (Cs-137 ,D∘=20mR/hr) gamma radioactive point source was positioned in different places in front of cross section of the cemented metallic pieces and gamma dose rates were measured on the outer side of the drum as in fig.(1). Readings showed good efficiency of the cement matrix to decrease the dose rate of (Cs-137) gamma radiation lower to acceptable values and with waste acceptance criteria and regulations. Fig (2 ) and table below show the dose rate measurement system and the variation of dose rate and attenuation coefficient in terms of dose rate with cement matrix thickness along cemented waste cross section or hypothetically along the Radwaste container diameter.

Speaker: Ms Sabeeha AL-TIMMIMI (Arabic-IRAQI)

09:00 DECOMMISSIONING AND ENVIRONMENTAL REMEDIATION PLANNING FOR NIGERIA RESEARCH REACTOR-1 (NIRR-1)

Centre for Energy Research and Training (CERT), operates a Research Reactor codenamed NIgeria Research Reactor-1 (NIRR-1). The reactor was commissioned February, 3 2004. The designed and operation of the NIRR-facility are based on both national and international regulations and safety standards. Thus at end of its lifetime, the facility shall be decommissioned so that the site will become safe for other use(s) and all items removed are secured. This paper presents overview of decommissioning and remediation plan of the facility with respect to organizational, technical, safety, and security as well as the legal and institutional framework.

Speaker: Dr Muhammad Auwal Musa (Centre for Energy Research and Training (CERT), Ahmadu Bello University, Zaria-Nigeria)

Paper DECOMMISSIONING_PLANNING_FOR_NIRR-1.docx

09:00 Decommissioning Experience in the IEA-R1 Research Reactor

ABSTRACT The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water, using graphite and beryllium reflectors. The reactor is located in the Institute of Energy and Nuclear Research (IPEN-CNEN / SP), in São Paulo, Brazil. The first criticality of the reactor was obtained on 16 September, 1957. Since the beginning of the reactor operation, the facilities have undergone a decommissioning process on two occasions: one in 1978, when the lining of the pool (tiles) was replaced by stainless steel, then, in the end of 2013, when parts of the cooling primary circuit were substituted. This paper describes these two events. 1. INTRODUCTION During the IEA-R1 reactor operating time, changes were made at the facility, enabling safer operation as well as longer lifetime of the reactor use. These changes began by doubling the cooling system, changing the control desk, the pool lining and part of the primary cooling circuit. The last two items allowed us to gain experience in decommissioning nuclear facilities. 2. POOL LINING CHANGE coating e lining são revestimento; verifique se tem uso especial para reator ou use os dois, para evitar repetição After almost twenty years since the first criticality of the reactor occurred, water leak from the primary refrigerating circuit and through the tiles that lined the pool of the reactor were identified by the reactor operation personnel. As it would be unfeasible to replace only some tiles, since the identification of each piece would be impossible, the total exchange of the coating was decided, based on the experience gained in other reactors and stainless steel was the material chosen. After a great load of work and effort, the complete removal of the original tiles and subsequent replacement by a coating of stainless steel plates was carried out. (Fig.1) The irradiated or contaminated material removed from the reactor pool was dry stored in drums and sent to the Radioactive Waste Laboratory (LRR), located at IPEN-CNEN / SP. Fig. 1 3. REPLACEMENT OF PART OF THE COOLING PRIMARY CIRCUIT In July 2013, a high degree of corrosion on the pipe supports of the primary circuit of the reactor cooling system was verified. There was, also, corrosion in the flange bolts. (Fig. 2) In principle, corrective actions were taken to replace the pipe supports and flange bolts. New supports were designed and replaced those which were old; however, when switching the screws it was necessary to use a blowtorch and hard strength, which caused the appearance of microcracks in the welding of the flanges. The problem was discussed in the reactor internal security committee together with the Nuclear Engineering Center team. In view of the situation, the partial exchange of the piping, improved shelter and re-sizing of the shield showed to be necessary. (Fig. 3) All piping withdrawn from the section affected was internally contaminated. As the extension of the pipe stretches was long, the pipeline was cut into pieces of 50 cm; after decontamination, some tubes were discarded as scrap and those, whose decontamination was not possible, were transferred to the Radioactive Waste Laboratory (LRR), located at IPEN-CNEN / SP. 4. CONCLUSION The reactor upgrade that was carried out and it is still ongoing allow us to have a projection of operation over the next 10 years. The procedures allowed the reactor operating staff to gain experience in decommissioning nuclear installation and equipment. It is very important to discuss the possibility of the reactor shutdown in the future and the need to have a decommissioning plan, which should offer the possible options as well as the skilled labor and the necessary funds required. 5. REFERENCES Frajndlich , R., IEA-R1 RESEARCH REACTOR: OPERATIONAL LIFE EXTENSION AND CONSIDERATIONS REGARDING FUTURE DECOMMISSIONING. International Nuclear Atlantic Conference - INAC 2009

Speaker: Mr Julio Benedito Marin Tondin (IPEN-CNEN/SP)

09:00 DECOMMISSIONING HAMAOKA NPS UNITS 1 &2 WITH ASPECT OF RADIOACTIVE PROPERTIES

1. The subject to realize decommissioning plan The most essential subject to realize decommissioning plan is to grasp the radioactive properties (Radioactive inventory) of the target installation. Based on the radioactive properties, Decommissioning Plan (including dismantling method, reduction of radiation exposure, safety evaluation, waste disposal plan, etc.) shall be realized, optimized and carried out. And radioactive waste shall be disposed with safety manner. Especially, reactor and inside core internals are difficult to handle them by direct handling for radioactive characterization by radiochemical measurement etc. Theoretical calculation is a typical evaluation technique especially for difficult-to-measure nuclides which are contained in activated wastes such as core internal metal. It is essential to improve evaluation method as close to the real value for the scientific rational waste disposal. In order to such that, the project below was established in order to improve calculation precision by comparison between calculation method of neutron induced activation and radiochemical measurement method, and by verification of evaluation method. 2. Radioactive Characterization for Hamaoka Unit-1&2 Hamaoka in Unit 1, core sampling and radiochemical analysis is carried out to verify the theoretical accuracy targeting "Reactor Vessel", "Inside Core internals" and "Containment Vessel Concrete Structures" since 2009. Sampling of "Reactor Vessel" and "Inside Core internals" has already been completed. Radiochemical analysis of radionuclide and chemical analysis of precursor element are conducted at the Hot Laboratory set in Hamaoka NPS. Verification and Adjustment of Calculation Method is planned to be completed by 2018. The same project is also planned targeting at Hamaoka Unit 2. Radioactive characterization of neutron induced activated structure is carried out by Chubu EPC jointly collaborated with EPRI which has a wide range of knowledge about decommissioning and activation calculation. 3. The objective Dismantling method, waste treatment and disposal method to be optimized by improving activation calculation accuracy. Development of sampling equipments and Hot Laboratory to be set up to survey radioactivity characteristics of actual material. It seems that the above Chubu Project will be internationally beneficial because there is few such a study example.

Speakers: Ms Hanae Kanzaki (chubu electric power Co.,Inc), Mr Seiji Komatsuki (Chubu Electric Power Co.,Inc.), Mr motonori nakagami (chubu electric power Co.,Inc), Mr yasuhiro Nakada (chubu electric power Co.,inc)

09:00 Decommissioning of radioisotope thermal generators (RTGs) in the Republic of Tajikistan

The Republic of Tajikistan, as well as other republics of the former USSR, used radioisotope thermoelectric generators (RTGs) as electricity power source for autonomous hydro- and weather-navigation equipment, located in hard-to-reach mountainous areas. Radionuclide heating sources (RHS) in all RTGs are made on the basis of Strontium-90 (half-life 29.1 years). Depending on type, RTGs may contain 185 to 12.950 TBq. The total activity of all produced RTGs taking into account daughter radionuclide – Yttrium-90 is about 3.7 PBq. RTGs were under permanent control in the former USSR. However, after the disintegration of the USSR, hundreds of these small facilities, equipped with powerful sources, have been left without control. The radioactive substance contained within them may easily be used as a source for a radiation dispersion device (RDD). Using the Strontium-90 (as a material for a common bomb, this radioactive substance can be dispersed after an explosion). Having detonated one of these “dirty bombs”, a terrorist could contaminate entire parts of towns. The only organization in the Russian Federation that develops, produces, makes general overhauls, modernization and prolongation of RTGs is the All-Russian Research Institute of Technical Physics and Automatization in Moscow. This Institute supplies RTGs to various ministries, bodies and enterprises. The main customers of the Institute are the Ministry of Defense, the Ministry of Transport, Goskomgidromet and the Ministry of Geology. According to unofficial data, there were 15 RTGs installed in the Republic of Tajikistan during the Soviet Union by Tajikgidromet (Tajikistan’s hydremeteorological service). After expiration of their operation life, most of these RTGs were disassembled and shipped back to the Soviet Union. Control over some of the RTGs in the Republic of Tajikistan was lost during the period of civil war. The employees of the Ministry of Extreme Situations and Civil Defense of the Republic of Tajikistan (MESCD) accidentally detected an emergency situation in the territory of a coal storehouse of Tajikgidromet in Dushanbe. This emergency situation was caused by the loss of one of four RTGs. The dose rate at the distance of one meter from the source was 180 microsieverts per hour. The procedure for handling RTGs found in the Republic of Tajikistan is planned to consist of four stages: •Stage 1. Perform an investigation and search mountainous areas by specialists of MESCD and Nuclear and Radiation Safety Agency (Regulatory Authority) of the Academy of Sciences of the Republic of Tajikistan (NRSA). Receive expert assistance for RTGs from specialists from Minatom of the Russian Federation and IAEA experts. Assistance should focus on how to store these RTGs in TRWRS and to determine their transportation requirements; •Stage 2. Transportation of RTGs from TRWRS by special transport to an interim platform and area of Tajik railways for loading to special carriage; •Stage 3. Transportation of RTGs in a special carriage by train to the Russian Federation for dismantling and extraction of the radioisotope sources. Radioactive sources will be placed in special transport containers and placed in special carriages; •Stage 4. Transportation of high-level radioactive sources, extracted from RTGs, in transport containers by special railroad car to State Enterprise “Mayak” (Ozyorsk city, Chelyabinskaya oblast) for disposal

Speaker: Prof. Ulmas Mirsaidov (Nuclear and Radiation Safety Agency)

09:00 DECOMMISSIONING STRATEGY FOR CERAMIC MELTER IN NUCLEAR WASTE VITRIFICATION OF HIGH LEVEL WASTE IN INDIA.

Abstract : Vitrification of high level liquid waste (HLW) by single step Joule heated ceramic melter (JHCM) based technology was developed in India in mid nineties to meet the growing demand of HLW management in near future.With the enhancement of Indian atomic energy programme substantial increase in HLW generation is expected in integrated reprocessing and vitrification plant under construction at Tarapur.Joule Heated Ceramic Melter (JHCM) technology has been adopted for industrial scale vitrification of high level liquid waste at Tarapur and Kalpakkam and also the reference melter technology planned to be adopted in the upcoming integrated reprocessing and vitrification plant at Tarapur. As a technology demonstrator a radioactive demonstration plant viz. Advanced Vitrification System (AVS) was designed, installed, commissioned and under operation in process cells of the Solid Storage Surveillance Facility (SSSF), Tarapur for treatment of HLW presently stored at Tarapur site. The AVS plant was successfully hot commissioned and around 175 Cu.M of high level waste was successfully vitrified and stored in interim storage facility of SSSF,Tarapur in the first phase of operation which lasted around three years, in the first vitrification cell ( AVS-1) of SSSF.While the second cell was being equipped with a new melter ( AVS-2), the melter operation in the first cell was discontinued and efforts were initiated to remotely remove the melter by cutting/grabbing/shearing the metal/refractory pieces. As the melter was placed in a retrofitted cell available in SSSF facility, it was felt that a mock-up facility must be created to test different tools and gadgets to test their life, endurance and efficacy in real situation while actual cutting and removal operation will be initiated later on. Key Steps of dismantling the AVS-1 melter were identified as follows: a) Removing removable top accessories such as thermocouples, thermocouple guide pipes, plenum heaters, and feed (frit & HLW) lines and off gas jumper. b) Cutting & removal of Melter top nozzles & inserts of thermowells, density probe, and level probe. c) Cutting & removal of Melter top plate. d) Cutting of inconel side electrodes. e) Breaking/Removal/ sizing of fiber wool, fiber board, bubble alumina, top refractory, side refractory, backup insulation, glass, inside pieces of side electrodes, bottom refractory, insulation block and inside piece of freeze valve electrodes. f) Removal of melter casing. g) Sizing & Categorization h) Packaging inside drums, canisters/over packs depends upon activity level. i) De-dusting and de-contamination of cell Considerable efforts were made to create and operate a mock-up facility where different cutting tools and gadgets for remotely cutting and removal of stainless steel plates, structural materials, ceramic parts and components of the melter, inconel plates and pipe were tried and the life cycle of each tools were tested and evaluated. The actual operation of decommissioning of AVS-1 melter was initiated and considerable progress was made in the first few months in removing almost 70 % of the components. The lesson learnt during the process was absorbed and the experience was taken into account while designing the ceramic melter for subsequent upcoming projects. It was also felt that for faster and easy decommissioning, the design of the glass contact refractory ceramic blocks plays a crucial role. The refractory block design was modified to take care of this aspect in future design. This paper describes the decommissioning experience of the first ceramic melter of Advanced Vitrification System (AVS ) ,Tarapur ,difficulties , lessons learnt and modifications incorporated in the design of future plants. References: Demonie, M., P. Luycx, M. Snoeckx and L. Baeten, 1994, “Experience Gained with the Dismantling of Large Components of the Pamela Vitrification Plant”, Belgoprocess, Dessel Belgium Brooks, R., 1995, “Slurry Fed Ceramic Melter Disassembly Report”, CMH032493, West Valley Nuclear Services Co., Inc., West Valley, New York. Evans, J., M. W. Noakes, and K. Kwok, 1995, “An Assessment of Tooling Requirements for a Dual-Arm Manipulator System Performing Decontamination and Dismantlement of Nuclear Facilities”, ORNL Internal Report, Oak Ridge, Tennessee.

Speaker: Mr Kalyan Banerjee (Bhabha Atomic Research Centre,Department of Atomic Energy,Government of India)

09:00 Decommissioning Strategy For Liquid Low-Level Radioactive Waste Surface Storage Water Reservoir

The paper discusses long-term radiation and radioecological safety assurance and justification for the Techa Cascade of water reservoirs (TCR) at PA “Mayak” created in 1951-1964 to store large volumes of LRW. Since the late 1990s, TCR operation has been associated with high radiation and radioecological risks due to the large size of the facility (TCR area exceeds 50 km2), the total accumulated activity (over 360 mln m3 of LRW with a total activity of some 5∙1015 Bq has been accumulated to date) and strong dependence of precipitation on TCR state. As the result, the facility established under the first stage of the river Techa remediation program started to pose a threat as only limited solutions providing effective water level control could be implemented. Unlike other industrial water reservoirs, TCR conservation involving soil cover placement, contaminated water retrieval or treatment is considered to be basically impossible. The estimated total cost of such efforts would have exceeded 200 bln RUB, whereas the estimated maximum damage worth less than 50 bln RUB. In the early 2000s, a series of urgent measures designed to improve the TCR water balance and hydraulic engineering structures stability was implemented due to the increase in regional water level and a real risk of TCR overflow. Relevant engineering efforts have been carried out in parallel with the development of calculation and monitoring tools necessary for effective management and justification of TCR long-term safety. By now, high population and environmental risks associated with the TCR state have been significantly reduced. In the last 5-8 years, a relatively stable water level enabled to implement a large-scale cross-disciplinary R&D program aimed to justify a set of efforts required to achieve the ultimate solution to TCR challenges presented in a form of a strategic master-plan. The paper presents the main results of this work, in particular those associated with:  identification of TCR end-state, justification of main strategies aimed at achieving it, as well as long-term work schedule and consolidated road maps with due regard to relevant top-priority tasks and boundary conditions;  development of calculation tools designed to forecast TCR long-term behaviour based on different engineering solutions and environmental conditions;  development of treatment technologies and installations for TCR water and LRW discharged into it;  addressing key regulatory issues.

Speaker: Mr Sergey Utkin (NUCLEAR SAFETY INSTITUTE OF THE RUSSIAN ACADEMY OF SCIENCES)

09:00 Design of Iraqi Management Database System, All Nuclear Decommissioning and Waste, Projects and Activity Tracer

Abstract Design and Implement a database system to regulate the work map of two Directorates (decommissioning and waste management) according to their exceptional work context and Task and Action plans concepts, the process to design a Database depends on two concepts; the first is take the beneficiary directorates each separately to define the data input and stand on the overlap between the input of each, the second is to take in consideration applying the Action and Task plan which materialize the best preservation and documentation, and to follow-up the activity process of each directorate. This project is Local Area Network LAN, window form application based on ASP.NET technique using C# programming language.

Speakers: Ms AZHAR ALSARRAY (29/11/2015), Mr SAAD ALMUKHTAR (29/11/2015)

09:00 DEVELOPMENT OF PRACTICAL GUDIANCE ON OCCUPATIONAL RADIATION PROTECTION FOR APPLICATION IN DECOMMISINIONG OF NUCLEAR FACILITIES

Abstract: Nuclear decommissioning is an increased industrial activity in nuclear fuel cycle facilities owing to the fact that some of the nuclear installations have outlived its usefulness and hence remained under permanent shutdown. These facilities require safe decommissioning and subsequent restoration of site to general use. In order to facilitate this activity, IAEA is rendering all possible help to Member States to strengthen their infrastructure to ensure safe decommissioning. It developed several safety standards to provide guidelines for safe and successful implementation of decommissioning. A project sponsored by European Commission UNDER Cooperation of Nuclear Safety (CNS) has been designed by IAEA to develop a practical guidance supporting the Occupational Radiation Protection (ORP) during decommissioning. This manuscript discusses about the issues related to occupational radiation protection aspects of decommissioning strategy that includes identification of hazards, risk assessment, fingerprinting of radionuclides responsible for causing exposures and optimization of radiation protection. Besides, this also focuses on protection measures for dismantlement of active components and radioactive waste management. The need for sound radiation protection programme which encompasses effective workplace monitoring and individual monitoring is also emphasized. 1. INTRODUCTION Occupational radiation protection plays important role entire nuclear fuel cycle activities. The sources of exposure in decommissioning phase are quite different from that of operational phase. Between the permanent shutdown and beginning of decommissioning activity there exists considerable time lag that causes short lived isotopes to decay down resulting in modification of type and magnitude of exposures. As a result of this, it is necessary to identify the dosimetrically significant radioisotopes that could probably enhance the exposures to workers. 2. METHDOS The objective of a decommissioning plan is to remove the radiological and non-radiological hazards associated with the operation of a nuclear facility. According to the Basic Safety Standards [1], arrangements must be made for the assessment of the occupational exposure of staff, on the basis of individual monitoring where appropriate, including arrangements with appropriate dosimetry services under an adequate quality assurance programme. Towards this, fingerprinting the radioisotopes that causes exposures needs to be established. Among methods available for this, the operational history of plant provides useful information. Besides, the computer modelling [2] could also be used to ascertain the sources of exposures which again need to be validated with experimental measurements. 3. RESULTS IAEA has developed a TECDOC, which provides the practical guidance related to RP aspects of decommissioning. It outlines the need for strong radiation protection measures and suggests methodology to establish the radiation protection programme considering the varied degree of risks associated. Prior to decommissioning, a thorough assessment of radiological and non-radiological hazards and their corresponding risks should be conducted, with continuous re-assessment throughout the execution of decommissioning activities. A systematic methodology has been proposed for risk assessment process. There are two predominant hazards present in work environment i.e. radiological and non-radiological risks. With the removal of fuel elements, the risk of exposure has appreciably reduced compared to regular operations. However, numbers of industrial hazard are expected to be more in decommissioning stage compared to operational phase. These hazards have potential to influence the exposures from radiological hazards. In view of this, the precise determination of the hazards is essential. This involves identification hazards encountered by workers in work environment. Hazards identification results in risk estimation which in turn helps to make a sound risk assessment. Based on the risk assessment ORP is implemented. Decommissioning of a nuclear installation brings about changes that may impact the prevailing safety culture. There can be a perception that when moving from routine operations to decommissioning, the importance of radiation safety is reduced if the fuel is removed from the reactor or site. This could adversely impact the safety culture and is incorrect as, despite removal of the fuel, the level of risk to workers is not necessarily reduced. High radiation levels may make deferred dismantling a more appropriate strategy because radioactive decay may allow radiation levels to decrease over time. However, there are limitations with respect to radionuclide composition and dismantling techniques. For example, if radiological relevant nuclides with long half-life are occurring (e.g.241Am or 90Sr), the decay of easily measurable gamma-emitters with shorter half-lives (e.g.60Co and 137Cs) might lead to a nuclide composition which is difficult to measure and radio logically more challenging. Furthermore, expected dose reduction of workers may not be achieved because remote dismantling is replaced by manual dismantling. When no benefits from radioactive decay or changes in radiological conditions to unfavourable nuclide compositions are expected, immediate dismantling is the preferred strategy. 4. CONCLUSIONS Decommissioning of nuclear facilities must be accomplished taking into account of radiological and non-radiological hazards. The optimization of protection and safety must be ensured. REFERENCES [1]. Basic Safety Standards, General Safety Requirements Part 3 (2014).

Speaker: Dr Suriya Murthy Nagamani (International Atomic Energy Agency)

09:00 ESTIMATION OF THE RADIATION DOSES DURING THE DISMANTLING OF THE EQUIPMENT IN BUILDING 117/1 AT THE IGNALINA NPP

The Ignalina nuclear power plant (NPP) was the only NPP in Baltic States, build in north east Lithuania near the border of Belarus. The Ignalina NPP was operating two world’s largest and most advanced RBMK-1500 design reactors (electrical capacity – 1500 MW, thermal capacity – 4800 MW). It supplied about 70% of Lithuania’s national electricity demand. In line with accession to the European Union treaty commitments, the Ignalina NPP was closed: Unit 1 was shut down at the end of 2004, and Unit 2 was shutdown at the end of 2009. Since 1 January 2010, decommissioning has become the major Ignalina NPP activity. The auxiliary plant systems can now be progressively dismantled. The first area to undergo dismantling was the Emergency Core Cooling System (ECCS) equipment of Unit 1 which is in Building 117/1. Building 117/1 is located close to Unit 1 of Ignalina NPP reactor building. It is a rather big building (the volume is 13748 m3) with more than 30 rooms located at -3.6–7.2 m level. The main part of the Ignalina NPP Unit 1 Emergency Core Cooling System (ECCS) is located in this building. The main systems which are located in Building 117/1 are the following: • Emergency Core Cooling System (ECCS); • Helium storage facility; • Nitrogen supply system. During the planning of D&D activities in Building 117/1, the overall dismantling activities were segregated into ten smaller activities. The first one is the preparatory activity, which includes of civil works, initial dismantling and modification or installation ventilation systems, installation cranes, preparation of temporary waste storage area, etc. After the preparatory activity, the dismantling activity can be started during which components such as ECCS pressured tanks (PT), large diameter pipes, valves are dismantled. According to the national law, during D&D activities, it is necessary to minimise the amount of radioactive waste, and therefore it is necessary to perform the decontamination activity. In order to perform successful decontamination of the contaminated internal large diameter pipe surface, it is necessary to halve these pipes, and therefore the third activity was defined as pipe halving. To achieve free release (FR) level of the components, the dry decontamination method (manual vacuum blasting technique) was selected during the analysis of D&D strategy options. During initial dismantling, dismantling, pipe halving and decontamination, the components are transported from origination area to a temporary storage area, then from the temporary storage area to a decontamination area, etc. These activities were called ‘handling’. Waste sorted by component type, contamination level (exempt waste, very low level waste, low level waste) was loaded into an appropriate container and transported to the final destination. In parallel with all described activities inside Building 117/1, the activities of management and radiological measurement and characterization of the components were performed. After dismantling of the components and transportation of all waste from the building, it is necessary to perform the close out activity, which consist of final monitoring of the building, decontamination and disassemble of D&D equipment and dismantling of cranes, etc. For preparation of the equipment dismantling project in Building 117/1, DECRAD computer code [2] was used. DECRAD was developed at the Lithuanian Energy Institute (LEI) by Nuclear Engineering Laboratory. In principle the dismantling activity defines the total duration of the D&D activities. The duration of the decontamination, handling, loading packages, transportation activities are of course dependent on the duration of the dismantling activities. Preparatory and close out activities compose 30% and 22% of all the D&D activity duration respectively (see Figure 3). Duration of the radiological measurement activity depends on duration of the dismantling and the decontamination activities. The management activity is performed during the whole project. In the paper the information on the necessary personnel and collective doses received for different activities will be presented. The predicted duration of the project and collective doses to the personnel will be compared with the real data obtained after the dismantling of the equipment in Building 117/1. REFERENCES [1] Environment Impact Assessment Report, Ignalina NPP Building 117/1 Equipment Decontamination and Dismantling, VT Nuclear Services Ltd. and Lithuanian Energy Institute (Nuclear Engineering Laboratory), 2008. [2] The software Decrad validation report, TA-14-13.10. Lithuanian Energy Institute, Nuclear Engineering Laboratory (2010).

Speaker: Dr Audrius Simonis (Lithuanian energy institute)

Poster Poster_Audrius_Simonis.pdf

09:00 KIT Competence Center for Decommissioning – An Overview

The process of decommissioning a nuclear facility is the last step in closing the life cycle of a nuclear power plant. Due to political, economic and technological issues, this activity has been significantly increased worldwide, creating demand and opportunities for highly-skilled workers. Therefore, the education of adequate personnel, developing new technologies and improving existing techniques is of utmost importance. The Karlsruhe Institute of Technology (KIT), with its long tradition in nuclear research, already features a broad range of know-how in the decommissioning of nuclear facilities at its different institutes and organizational entities. In order to bundle this expertise, allow synergies within the research activities and intensify joint research, the “Competence Center for Decommissioning” has been established in February 2015 at KIT. This presentation will introduce this new entity and its major partners and give a short overview of current research activities.

Speaker: Mr Martin Brandauer (KIT)

09:00 LEGACY OF THREE DECADES OF FUEL CYCLE DEVELOPMENTS, DECOMMISSIONING ACTIVITIES ACCOMPLISHED AND CHALLENGES AT IPEN-CNEN/SP, BRAZIL

The legacy of some decades of nuclear fuel cycle development activities accomplished at IPEN-CNEN/SP as well as the main problems faced and the challenges for the decommissioning of the remaining facilities and environment restoration needs are discussed In this paper.

Speaker: Dr Paulo Ernesto Oliveira Lainetti (Nuclear and Energetic Research Institute - IPEN-CNEN/SP Brazilian Nuclear Energy Commission)

09:00 MANAGEMENT OF ALPHA CONTAMINATED UN-SERVICEABLE GLOVE BOXES.

Abstract: - India with it’s closed fuel cycle policy, like few others in the world, has it’s own share of legacy waste. Slow but steady accumulation of α-contaminated un-serviceable Glove Boxes (GBs) is a concern and demands solution for management of such radioactive waste. As a temporary measure these GBs, encased in individual secondary enclosures, were transported and stored at dedicated alpha storage facility. In 2014, a campaign for contact dismantling of these α contaminated GBs was planned. Under this campaign, a temporary facility for active contact dismantling was developed, designed & erected in the existing alpha storage facility and six (6) GBs stored were successfully dismantled and disposed. This report talks about the practices being followed in the nuclear industry around the world, the inactive dismantling trials in BARC and finally about the campaign for active contact dismantling of alpha contaminated GBs. 1. Introduction: Extensive dismantling trials were carried out on dummy GBs using various portable cutting tools in prototype setups based on remote, contact and cutting through glove port modes. Cutting through glove port appeared the best among the three options considering many important criteria. But a decision was taken to go for a contact dismantling campaign in a temporary set up at alpha storage facility due to several constraints and considerations; one of the reasons among them was to create spare storage space for receiving such waste already accumulated in the laboratories. 2. Common International practices on management of GBs: 3. Contact Dismantling Facility at AWTSF: A primary containment (PC), an airlock, secondary enclosure (SE) with local HEPA filtration was erected. The PC and the airlock was made of PVC sheet. The SE was made of Perspex and metal sheets. Existing negative pressures system, ventilation system was augmented for the dismantling facility. All necessary safety systems and radiological monitoring systems were installed. The dismantling of GBs was carried out manually by operators wearing plastic suit and fresh air line connection. Detailed internal swipe samples of each and every GB were carried out in PC by removing their glass panels. GBs were decontaminated prior to cutting and the cut pieces were bagged out and stored in the drums. The same setup was reused for dismantling of all the six GBs by proper DC of the airlock and PC. 4. Important findings of the campaign: • The cut pieces and secondary waste generated were stored in drums. • Glass panels of GBs were stored in special containers. • The air activity was found well within the limits and no spread of contamination outside SE. The campaign was safely executed by adhering to the best radio-logical practices. • Internal dose assessment of the personnel involved was found below detectable limit. 5. Characterization & Disposal of waste:- A methodology was formulated to estimate total activity content based on various swipe survey data. It was observed that the cut pieces of the GBs qualify for disposal to NSDF after cementation. Accordingly the drums filled with the cut pieces were disposed after cementation as CAT-I non alpha solid waste. 6. Conclusion:- • The campaign, apart from the experience of a pioneering venture, has generated valuable data not only on the radiological aspects like generation of air activity, level of contamination but also on the mechanical issues like tools & handling systems. • It has been observed that there is substantial volume reduction and the same can be further enhanced by adapting to scheme of cutting through glove port and reusable enclosures. • Dismantling of GBs allow detailed assaying and through DC which helps in segregation and disposal of the waste, besides volume reduction. 7. References • State of Art Technology for Decontamination and Dismantling of Nuclear Facilities, Technical Report Series No. 395, International Atomic Energy Agency, Vienna 1999. • "Design Innovation For The Management Of Alpha Contaminated Unserviceable Glove Boxes", Devendra Sandhanshive, BARC, Trombay, et al, Proceedings of the ASME 2013 15th International Conference on Environmental Remediation and Radioactive Waste Management Sep 8-12, 2013, Brussels, Belgium, [2013]

Speaker: Mr Devendra Sandhanshive (Bhabha Atomic Research Centre (BARC), Trombay, Mumbai, India)

09:00 Monte Carlo techniques in Radiological Characterization for Reactor Decommissioning

Radiological characterization of shut down nuclear reactors is one of the most important and basic steps because it can directly affect the whole approach to decommissioning. For this purpose Monte Carlo techniques can generate a specific activity for each radionuclide present in the material in a particular region of interest whether it results from radioactive releases from the fuel, or it is activated product which occurred during normal operation or unplanned events. Monte Carlo techniques are almost irreplaceable in the situation when is necessary to have a realistic assessment of the activities of hard-to-detect radionuclides in a mixture of various radioactive substances when scaling factor approach can be used. In this work MCNP geometry model of semiconductor Si detector, Ge detector and GM hand probe are used for radiological characterization during decommissioning of shut down research reactor RA at the Vinča Institute.

Speaker: Dr selena grujic (Faculty of technical sciences, PC Nuclear Facilities of Serbia)

09:00 PREPARATION OF DECOMMISSIONING PLAN OF BANDUNG TRIGA RESEARCH REACTOR

Abstract: In accordance with the decision of the head of the regulatory body of Indonesia related to the decommissioning of nuclear reactors, the operator must prepare a decommissioning plan. To achieve these objectives, in last two years, the research related to the environmental assessment around the reactor building, decommissioning cost estimates, radioactive waste management and the estimation of radioactive material inventory at the beginning of shutdown conditions have been done. To support the implementation of these activities needed an expert advice, comparative study to the type of reactor that has been carrying out decommissioning and training related to decommissioning activities. These activities are part of the planned activities in the TC program in the period 2014 to 2015. Recently, the decommissioning plan of Bandung TRIGA research reactor is in the stage of completion and is expected to be completed later this year. This paper discusses the national policy and strategy related to Bandung TRIGA research reactor and the activities undertaken in preparation of the decommissioning plan of Bandung TRIGA research reactor 1. INTRODUCTION Currently, there are three research reactors operating in Indonesia. Those are Bandung TRIGA research reactor(2000 kW), Kartini research reactor (250 kW) and Siwabessy multipurpose reactor (30MW). These reactors are operated by the National Nuclear Energy Agency (BATAN). The Bandung TRIGA reactor reached first criticality in year 1964. In 2001, the TRIGA 2000 reactor received the license operating from regulatory body until 2016. Since September 2011, the reactor operating license suspended by the regulatory body because there FFCRs have burn-up more than 50% and program retrofitting has not been done. The government of Indonesia (BATAN) expects Bandung TRIGA reactor can be operated more safely and strongly supports the continued operation of Bandung TRIGA reactor. For this purpose, BATAN upper management has promoted marketing team and technical team for the Bandung TRIGA research reactor. The marketing team prepared the strategic plan in order to gather all potential stakeholders (academics, industry, medical application, environment surveillance). Meanwhile, the technical team has assessed the different technical options and related economical costs for bringing the Bandung TRIGA research reactor back in operation. Recently, a strategy plan of Bandung TRIGA research reactor for future activities has prepared to assure the strategy plan goals could be achieved. We planned several options to continue operating the reactors again such as procurement of fuel element standard TRIGA, utilizing existing TRIGA fuel and create new control rods without fuel elements as a substitute FFCR (fuel follower control rods) and additionally decommissioning plan should be prepared. In the implementation of this program, some experts from various countries, coordinated by the IAEA, has come to give his experience to us. Recently, the decommissioning plan of Bandung TRIGA research reactor is in the stage of completion and is expected to be completed later this year 2. RESULTS The objective technical assessment conducted in Bandung TRIGA reactor to evaluate whether Bandung TRIGA reactor feasible to be operated or conducted for decommissioning. Accordingly, it is necessary by BATAN leaders to conduct in- depth study of Bandung TRIGA reactor, as consideration for determining the continued operation of Bandung TRIGA reactor. Results of the study: Option-1. Bandung TRIGA reactor resumed operations with the purchase of fresh standard TRIGA fuel elements including purchasing new FFCR. It seems this option cannot be implemented. Possibility standard TRIGA fuel elements are not produced anymore. Option-2. Bandung TRIGA reactor to continue operating for 1000 kW by using standard TRIGA fuel elements are still exist and replace FFCR with control rod elements without fuel follower. Option -3. To convert standard Bandung TRIGA reactor fuel elements into fuel plate type, such as the MTR reactor fuel. The problems to be faced in the third option is to design, create and install the core. Option-4. When the above three options are not feasible, then the option is implementation the decommissioning program for the Bandung TRIGA reactor. However, operated or not the Bandung TRIGA reactor decommissioning plan should remain in progress. Recently, the decommissioning plan of Bandung TRIGA research reactor is in the stage of completion and is expected to be completed later this year. 3. CONCLUSIONS 1. The national policy and strategy related to Bandung TRIGA research reactor has been made and strategic plan of Bandung TRIGA research reactor has been proposed. 2. The decommissioning plan of Bandung TRIGA research reactor is in the stage of completion and is expected to be completed later this year. REFERENCES [1] ANONYMOUS, BAPETEN CHAIRMAN REGULATION (BCR) No.8/2008 on the Safety Provision for Research Reactor Ageing Management, Jakarta, 2008. [2] ANONYMOUS, BAPETEN CHAIRMAN REGULATION (BCR) No.4/2009 on Nuclear Reactor Decommissioning, Jakarta, 2009. [3] ANONYMOUS, Strategic Plan of Bandung TRIGA Research Reactor, BATAN, Jakarta, 2014. [4] ANONYMOUS, Preliminary Decommissioning Plan of Bandung TRIGA Research Reactor, BATAN, Jakarta, 2012.

Speaker: Dr Efrizon Umar (National Nuclear Energy Agency od Indonesia)

09:00 PREPARATION OF DECOMMISSIONING PROGRAM FOR NPP WITH SHARED OWNERSHIP

The paper presents the general overview of preparation, validation, review and steps of approval of Decommissioning Program (DP) for NPP with shared ownership. The Krško NPP is a nuclear power plant in Slovenia, in commercial operation since 1983. It is a 700 MW two loop PWR NPP designed and constructed by Westinghouse. In the time of the construction the power plant was constructed in Yugoslavia, which was a federal state consisted of several federal republics. The investment in the power plant was a joint venture undertaking of utilities from two federal republics, namely Slovenia and Croatia. Joint investment was by Slovenian and Croatian utilities in equal shares. Since 1991 Slovenia and Croatia are independent states, and since 2004 and 2013 both countries (Slovenia and Croatia respectively) also member states of the EU. NPP is situated in Slovenia and its operation is governed by Slovenian legislation and Slovenian Nuclear Safety Administration (SNSA) is the regulator body in the Slovenia. Currently the plant is owned by state-owned Slovenian and Croatian utilities (i.e. GEN energija d. o. o. and Hrvatska Elektroprivreda d. d., respectively). All legal and other issues regarding the NPP Krško were agreed between the countries by a bilateral Agreement between the Government of the Republic of Slovenia and the Government of the Republic of Croatia (hereinafter the contracting parties) on the Regulation of the Status and Other Legal Relations Regarding the Investment, Exploitation and Decommissioning of the Krško NPP (hereinafter the Agreement). The bilateral Agreement which entered into force in March 2003. According to the Agreement the Program of NPP Krško Decommissioning and Program for SF and LILW Disposal (hereinafter the Programs) should be approved before put into force by both contracting parties trough the Intergovernmental Commission (hereafter the Commission) supervising the implementation of the Agreement. The main purpose of the Programs is to estimate the overall expenses of the future decommissioning, radioactive waste and spent fuel management of the NPP Krško in order to collect sufficient financial assets in separate funds in Croatia and in Slovenia. The paper will give general overview and some organizational aspects of preparation the expert work for the preparation and approval of a decommissioning program for a NPP with shared ownership according to the Agreement.

Speaker: Mr Leon Kegel (ARAO-Agency for Radwaste Management,)

09:00 PRIMARY AND SECONDARY COOLANT TREATMENT ON THE DECOMMISSIONING OF THE BN-350 FAST REACTOR

Abstract: Decommissioning of the fast reactor BN-350 in Kazakhstan was started in 1999. The government of the Republic of Kazakhstan has decided by its decree to bring the BN-350 in a state of long-term safe storage (“SAFSTOR” condition) for 50 years, followed by the dismantling and disposal. The main objectives to be achieved in the transfer BN-350 in a state of long-term safe storage is to achieve nuclear and radiation safety and the acceptable level of industrial safety. This decision means minimizing work at the initial stage of the decommissioning due to the limited funding available to implement the full-scale decommissioning project. Decommissioning strategy for the sodium coolant includes the following steps [1]: 1. Cleaning from cesium isotopes; 2. Draining from the loops and the reactor vessel; 3. Processing into chemically passive product and placed for safe storage; 4. Cleaning of inner surfaces of the reactor vessel and loops from sodium residuals.

Speaker: Mr Alexander Blynskiy (Nuclear Technology Safety Center)

09:00 PROJECT APPROACH TOWARDS “RADIOACTIVE WASTE PROCESSING AND DECOMMISSIONING AT VINČA SITE, SERBIA/TC PROJECT SRB3004”

The project “Radioactive Waste Processing and Decommissioning at Vinča Site” is part of the VIND program. The overall objective is to improve the safety of storage of radioactive waste on the Vinča site. This implies to carry out a detailed characterization, sorting, repackaging and rearrangement of the existing radioactive waste (currently stored in the deteriorated H1 and H2 hangars) and to store this waste in the new H3 storage facility on-site, as well as dismantling one of the former storage facilities. In this paper the strategy, organization and status of the project is explained. INTRODUCTION In the mid-1950s two research reactors were established on the Vinča site: the RA reactor (research reactor for high power irradiation services of a variety of experiments) and the RB reactor (research reactor to provide criticality research data). In 2004, twenty years after shutting down the RA reactor, the Serbian Government decided to institute the VIND program (Vinča Institute Nuclear Decommissioning) in order to commence decommissioning activities for the RA reactor and associated nuclear facilities. The VIND program consists of three major projects, of which the first project has been completed: 1. Repackage and repatriate spent nuclear fuel to Russia; 2. Implement radioactive waste processing and storage capabilities on-site; 3. Decommissioning activities. The on-going project is related to these last two projects. The purpose of this paper is to describe how the Contractor will implement all project tasks in order to fulfill the project aims. PROJECT ORGANIZATION The project is defined within a contract between the IAEA (the Client), PC NFS (the Beneficiary) and the Consortium TECNUBEL-IRE ELiT (the Contractor), in which TECNUBEL acts as leader. All project management activities and expert services lie within the responsibility of this Consortium. Through on-the-job trainings and thorough follow-up, the experts of the Contractor train the workers of PC NFS in the correct execution of the project tasks (e.g. source conditioning, waste characterization…). STRATEGY In order to achieve the most efficient implementation of all project tasks, a methodology was elaborated in which several tasks were enabled to be executed in parallel. In this regard the project began with the refurbishment of the Waste Processing Facility (WPF), after completing the required project documentation. Through the upgrade of the WPF it becomes possible to use its functionalities for the treatment of radioactive waste coming from H1 and H2. In the WPF several zones will be demarcated to perform the main project activities: source conditioning, smoke detector dismantling, segregation and repackaging of drums… Sketch of the project site Pictures of the H1 hangar During this early stage of the project, focus lies on the licensing activities regarding the permission to start these (nuclear) activities, since these particular licenses influence the starting time of the most crucial project activities. Following the issuing of the necessary licenses, the Contractor will begin with the removal of waste items out of H2. After taking the necessary measures in accordance with the developed procedures (e.g. contamination measurements), the packages will be transported to the WPF where a pre-characterization will be executed. Subsequently drums will be brought into the ventilated tent, where they will be opened by trained workers and the waste will be sorted according to the waste type and in accordance with the Waste Acceptance Criteria. Sources will be brought into the designated room for source conditioning and smoke detectors will be entered in the designated room for smoke detector dismantling. Following the segregation/conditioning/dismantling activity, the filled drums and containers will be transported towards the location next to the WPF where a final characterization will be performed. Subsequently the packages will be transported to their storage location in the H3 facility. When a large part of H2 has been cleared of waste items, an area will be demarcated in this facility which will subsequently be characterized (and if required, decontaminated). This area will be used as a buffer for the waste from H1. The Contractor will thus at this phase in the project empty H1 in parallel with H2.This parallel approach enables a faster emptying of H1, which will thus be able to be characterized, decontaminated and dismantled at an earlier stage in the project, in accordance with the approved documentation. When all items have been removed as well out of H2 and it remains completely empty, a characterization campaign of this hangar will be performed. If required, decontamination activities will be performed. CONCLUSION The complexity of this project lies within the parallel execution of various tasks, requesting skilled project management and large expertise. At this moment (November 2015), works on-site will start. The Contractor’s proven experience in this type of projects enlarges its confidence to finalize the project tasks within the planned three years.

Speaker: Ms Bieke Bekaert (Tecnubel)

09:00 RADIOANALYTICAL SERVICES WITHIN DECOMMISSIONING OF OLD FACILITIES IN ÚJV ŘEŽ, A. S.

History of ÚJV Řež, a. s. began 1955 and covered activities like nuclear physics, chemistry, nuclear power, experiments at the research reactor and many other topics. Main issues addressed in ÚJV in the past decades were concentrated on research, development and services provided to the nuclear power plants operating VVER reactors, development of chemical technologies for fuel cycle and irradiation services to research and development in the industrial sector, agriculture, food processing and medicine. Decommissioning of old obsolete facilities began in 2003 and will be finished in 2016. Due to the lack of information about long-term unused facilities and heterogeneous waste huge radiological characterization of the facilities had to be provided. The characterization consisted of non-destructive and destructive procedures. For non-destructive assay our laboratory is equipped with an in-situ gamma spectrometry, a segmented gamma scanner, and a digital radiography system. Destructive assay is provided in the central analytical laboratory department with capability to provide all requested radiochemical analysis.

Speaker: Mr Karel Svoboda (ÚJV Řež, a. s., Fuel Cycle Chemistry and Waste Management Division)

09:00 RECLAMATION OF A UNDERGROUND STORAGE OF RADIACTIVE WASTE – PIT 7.1

1. INTRODUCTION The «Pit 7.1», known also as «The Monolith», is a reinforced concrete prismatic structure, composed by four squared section wells. The Monolith dimensions are: width 1.56m, length 5.81m, depth 6.45m (total 59m3). Near the monolith, a drain pumping system was realized to collect the groundwater. The Monolith was the result of a program of waste disposal in the ground, carried out in the’70s and ‘80s, that saw metallic drums disposed of in each of the wells (Figure 1). Such drums contained in particular: heads and foots of the fuel assembles, technological waste derived from hot cells decontamination, pool water purification system filters and resins for a total activity of 10TBq. All materials were generated by «Nuclear Tests» activities at the ITREC plant. After completion of the disposal, the Monolith was completely buried and a concrete slab was built on top of it, as commonly in use at that time. The disposal was seen as final and no recovery was planned or envisioned. However, the Ministerial Decree of July 2006 ordered the reclamation of Pit 7.1 as a preparatory activity of the overall dismantling of the ITREC Plant. 2. METHDOS The preparatory activities carried out are as follows: 1) Hydraulic isolation of the monolith by a series of secant foundation piles placed deep in the clay layer in order to insulate the monolith from groundwater; 2) Construction of a capping beam; 3) Construction of both a static and a dynamic containment composed by a steel structure to support a PVC tent and a ventilation system to ensure the necessary air pressure differential needed to prevent possible leakage of contamination; 4) Implementation of a radiological monitoring system composed by an environmental Gamma Monitoring system and α, β, ϒ monitoring system of the off-gas place on the extracion pipe of ventilation system. The Italian Control Authority (ISPRA) licensed the Operational Plan in 2013 to carry out both excavation activities aimed at uncovering the monolith and probing activities in order to ascertain the structural characteristics of monolith and the drums’ positioning. All the removed soil was placed in prismatic containers (1 or 3 m3) and radiologically characterized. To carry out the investigation about the mechanical property of the monolith, a partial sacrifice of the external layer made of plaster was needed. During such activities, a percolation of liquid from inside of the monolith occurred. The liquid was collected analyzed and found to be radioactive. Consequently, the soil which was interested by the percolation event was reclaimed and disposed of. The hypothesis of absence of free liquids, which was part of the project’s assumptions, revealed itself to be incorrect after the anomalous event. Therefore, indirect investigations were carried out to delineate the different materials (steel, concrete, water) present inside the monolith, such as those involving ultrasonic tests. 3. RESULTS The chromatographic map showed on the right gives indications about the nature of materials: 1)Positioning of drums; 2)The concrete and steel rebars, showing separation between the wells and boundary structures (brown color); 3)The possible presence of water (violet color) Such surveys included: 1) External visual surveys by partial scarifying of the plaster. The scarifying of plaster was the same activity which procured the leaking of contaminated ; windows were opened where the indirect investigations suggested the possible presence of liquid. 2) Internal visual surveys through drilling holes in the monolith to allow the insertion of cameras. In order to control possible liquid leakages, the holes have been drilled in the structure with a downward inclination and valves were inserted into the. These activities demonstrated an absence of liquid along the lines of the vertical cutting such as future activities of cutting may take place in safe conditions. 4. CONCLUSIONS The recovery of the Monolith is an ongoing process. Future steps include segmenting longitudinally the monolith by diamond wire, cutting the basement horizontally and placing each well in a shielded and waterproof quarterdeck. All these operations will be carried out inside a dynamic system of confinement. Eventually, the debris resulting by such activities will be moved and stored in a specific on-site storage facility known as Building 9.3.

Speaker: Mr Vincenzo Stigliano (Sogin S.p.A.)

09:00 TECHNICAL FEATURE OF APPLICATION OF SPO TECHNOLOGY IN INDUSTRIAL IMPLEMENTATION UNDER DECOMMISSIONING PROJECT OF FAST REACTORS

Since 2002, research reactor BR-10 (RR BR-10) moved into a decommissioning status. The main scientific and technical challenge for decommissioning project was to determine the most safety technology for conditioning radioactive waste (RW) of alkaline liquid-metal coolants (LMC). As a decision has been the technology of solid-phase oxidation (SPO) of alkaline metal by slag from copper smelting industry [1]. At the initial stage of development of SPO technology were experimentally designation parameters and technical modes of conditioning and establish the course of chemical interaction of alkaline metal with oxides which are present in the slag. But the use of the obtained results in decommissioning project of RR BR-10 was required to conduct a large-scale verification of SPO technology. As a result have been developed two alternative ways of mixing the components. One of them involves the discharge of hot slag into the molten alkaline metal (“upper dumping”), the second - injection of molten alkaline metal under the layer of slag (“bottom feed”). To test the mixing methods was carried out a large-scale verification of one-time conditioning for 50 liters of alkaline metal at the experimental rigs, layout of which are shown in Figure 1. Figure 1. Mixing schemes of SPO technology. Thus, the large-scale verification was shown that the “bottom feed” has a number of advantages in terms of safety process control, efficiency and reliability of the design, quality of final product which are suitable for long term storage [2]. As a result, the “bottom feed” was designation for the implementation of decommissioning project of RR BR-10. Technical solutions that were worked out at the experimental rigs were introduced in the principle of operation of the Testing Ground under the decommissioning project of RR BR-10. A separate part of the Testing Ground is a unit Magma which is intended one-time conditioning for 60 liters of RW of alkaline LMC. Works which are concerned with industrial application of SPO technology will begin on unit Magma in 2016. Recently, the technical conditions are developed for the implementation SPO technology in relation to energy power reactors and reactors-prototypes. The result of development shows that the increase one-time conditioning is necessary to upgrade some components and equipment for more safety and effectively. REFERENCES [1] SMYKOV, V.B., KRUCHKOV, E.A., BAGDASAROV, J.E., et al., "Treatment of radioactive waste alkaline liquid metal coolants of research reactor BR-10"  Nuclear and Environmental Safety (international journal). – 2011. – N. 3 – P. 105. [2] INTERNATIONAL ATOMIC ENERGY AGENCY, "Treatment of Residual Sodium and Sodium Potassium from Fast Reactors" (Review of Recent Accomplishments. Challenges and Technologies), IAEA-TECDOC-1769, IAEA, Vienna, Austria. – 2015.

Speaker: Mr Kirill Butov (Rosatom, Russian Federation)

09:00 THE ASSESSMENT OF EXPOSURE DOSE TO THE RESIDENTS AFTER DECOMMISSIONING OF KORI NUCLEAR POWER PLANT UNIT 1 SITE WITH RESIDUAL RADIOACTIVITY ANALYSIS USING RESRAD CODE

In 2015, Korean government decided to decommission Kori nuclear power plant unit 1 site. In this study, we tried to assess the exposure dose to resident after decommissioning of Kori nuclear power plant unit 1 site with residual radioactivity analysis and the possibility of opening site to public. For this approach, we decided radionuclides in decommissioned site and the concentration of them, and the geometrical shape and volume of this contaminated site. We considered a variety of exposure paths from contaminants to human body. The main purpose of decommissioning is to decontaminate the areas, which are exposed to radioactivity, down to a normal level in order to release them area back to a natural environment. To evaluate the possible dose exposure which may be caused by a trace of radionuclide in the decommissioned site, we considered several cases of each involved exposed paths. We used RESRAD code for the assessment of residual radioactivity over time.[1] We considered the paths of external exposure, respiration and intake to human body, and liquid and gas radionuclide in Kori nuclear power plant unit 1 site. We applied weather conditions near Kori nuclear power plant unit 1 for recent 10 years. The results were huge different 1970’s, when built Kori nuclear power plant unit 1 site. Additionally, we improved reliability of results by analyzing and comparing dose conversion factor ICRP60 and ICRP103. In conclusion, the result was exceeding more than 10 times of 2.5mSv/y, which is NRC regulation in US, in the early stage of decommissioning. However, it decreased down to the normal dose 40 years after the decommissioning. After government making a decision the method which immediate or delayed dismantling, we have to study effects on decommissioning procedure and reassess periods of opening site to public.

Speaker: Mr Seokyoung Ahn (Pusan National University)

09:00 The Importance of Engineering Design in Reducing Risks During Decommissioning Hazardous Installations

Abstract The main objectives of decommissioning are to place nuclear facilities, that have reached the end of their useful lives in such a condition that they pose no unacceptable risks to the public, to workers or to the environment, and to reuse facilities and sites for new purposes. for that, Attached particular importance to reducing risks for people. as a result appropriate consideration of health and safety During the design stage, which covers concept detailed design specification (drawings, calculations, specifications, etc). for maximum potential for reducing risks, by application of the principles of safer design.

Speaker: Dr nadia sirag (Egyptian)

09:00 THE PROGRESS IN ARMENIAN NPP DECOMMISSIONING PLANNING

Abstract: The decommissioning issues had not been considered for ANPP at the design stage. The planning activities for ANPP decommissioning started in 2005. This paper describes implemented and on-going activities for ANPP decommissioning, including implementation of the pilot dismantling project on two systems of shut down Unit 1. 1. INTRODUCTION The ANPP consists of two units of the WWER/440/270 model, a modified version of the WWER/440/230 in view of special seismic considerations. Unit 1 started its operation in 1976 and Unit 2 in 1980. Both units were shut down shortly after the 1988 Spitak earthquake. The Unit 2 restarted operation in 1995. Design life-time of Unit 2 expires in 2016. Unit 1 is in a long-term shut down mode. Decision is taken on life extension of Unit 2 operation. The safety of ANPP is regulated by the Armenian Nuclear Regulatory Agency (ANRA). 2. DECOMMISSIONING PLANNING 2.1 Decommissioning Strategy and Plan The strategy for ANPP decommissioning was developed within EU TACIS Project in 2006 and adopted by the Government of Armenia in 2007. The selection process was based on the IAEA recommendations [1,2]. The strategy is called the Sequential Dismantling. In view of the Unit 2 lifetime extension the Strategy document will be revised. Initial decommissioning plan (IDP) was developed within TACIS On-site Assistance Project by UK Babcock Company (2010). The main purpose of the IDP was to initiate a ‘live’ document that will evolve as ANPP moves closer to the end of its operation. A Gap Analysis report on the IDP had been prepared (ANPP and Babcock experts) and identified a series of activities to address areas, where the IDP does not fully meet the IAEA recommendations [3]. Based on GP report the Action Plan was developed identifying the areas for improvement (i) develop the details of decommissioning strategy implementation; (ii) develop detailed cost estimate; (iii) develop Decommissioning Information System; (iv) develop Waste Management Strategy. 2.2 COST ESTIMATE The ANPP decommissioning cost was estimated by the Slovak subcontractor (NS and Decom Company). The cost estimate corresponds to the level of the IDP, i.e. budgetary estimate. Cost calculation methodology is based on the International Structure for Decommissioning Costing of Nuclear Installations [4]. Detailed WBS was proposed and related decommissioning schedule. Total cost is 545 M€. 2.3 ON-GOING DECOMMISSIONING ACTIVITIES Currently within EU Instrument for Nuclear Safety Cooperation Action Programme 2009 the decommissioning related project is implemented. Overall objective of the project is to develop the detailed ANPP decommissioning concept and selected licensing documents (within Part 1) and start pilot implementation of the concept and approach for Unit 1 selected systems (within Part 2). Specific objectives of Part 1 - (i) develop the concept at a high & intermediate level describing the major processes with their general interrelations – Process Map, and the conceptual ideas/ logic behind these processes – Process Model; (ii) develop licensing documents - SAR and EIA Report;(iii ) develop Decommissioning Information Management System; (iv) develop Decommissioning Waste Management Program. Specific objectives of Part 2 - applying the systematic approach prepare full set of D&D documents for a pilot dismantling project for two selected systems of the ANPP Unit 1 and conducting all necessary operations for the achievement of a complete D&D for these systems. Dismantling activities are in progress. 3. CONCLUSION The baseline planning, including the cost estimation for decommissioning, are important parts of the planning for ANPP decommissioning. The results of this phase are essential for safe implementation, a proper identifying the risks for future decommissioning, scheduling and the allocation of the financial resources. REFERENCES [1] INTERNATIONAL ATOMIC ENERGY AGENCY, Decommissioning of Facilities, IAEA Safety Standards Series, No. GSR, Part 6, IAEA, Vienna (2014). [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Decommissioning Strategies for Facilities Using Radioactive Material, IAEA Safety Report No.50, IAEA, Vienna 2007. [3] INTERNATIONAL ATOMIC ENERGY AGENCY, Standard Format and Content for Safety Related Decommissioning Documents, IAEA Safety Report Series, No. 45. [4] International Structure for Decommissioning Costing (ISDC), OECD/NEA, IAEA, 2012.

Speaker: Dr KARINE GHAZARYAN (ANPP CJSC)

09:00 TRANSITION FROM OPERATION TO DECOMMISSIONING OF HWRR IN CHINA

The paper aims to describe the activities conducted during transition between permanent shutdown and decommissioning of Heavy Water Research Reactor (HWRR) in China. HWRR is the first nuclear reactor in China. It went first criticality and was put in operation in 1958. It is a tank-type, heavy water cooled and moderated research reactor with 6 neutron beams and 1 thermal column. At the beginning of 1980s, it was modified to improve its performance. After modification, thermal power is increased from 7MW to 10MW, while maximum thermal neutron flux rate is increased from 1.2×1014 n/cm2.s to 2.6×1014 n/cm2.s. After 49 years operation, it was permanently shut down at the end of 2007. As a multi-purpose research reactor, HWRR has made great contributions to development of nuclear science and technology in China. After permanent shutdown, a project was approved and has been implemented. Reactor was defueled and spent fuel assemblies were moved to the storage pool for interim storage. Coolant in primary coolant system and secondary system was drained after reactor defueling. Heavy water was loaded into 200L barrels for storage. Spent fuel transport has been routinely conducted several times as it had been done many times during operation. Most spent fuel assemblies have been transported to the reprocessing plant except those discharged after final shutdown. Operational waste was cleaned up in the main reactor building and auxiliary buildings, including neutron guide tube of CNS and instruments and equipment of neutron beams and thermal column. As a result, the space of reactor hall was recovered for reconfiguration during decommissioning. Water containing K2CrO4 in the shielding tanks was drained and treated, so density of Cr6+ met relevant criteria and it can be drained as normal effluent. Radiological characterization survey of most structures, systems and components was conducted, so the overall radiological information was obtained. Of course, the reactor block is the key point. First of all, original composition of different structures was obtained by sampling and analysis. Then calculation, sampling and measurement were conducted respectively. But unfortunately, most internal components have not been sampled and measured, since accessibility to them is nearly possible. Supporting systems were modified, e.g. ventilation system, radiation monitoring system and entrance to the reactor hall. In the meantime, international cooperation on HWRR decommissioning was initiated and has been implemented step by step. Two IAEA TC projects have been successfully implemented. Furthermore, staff in different fields participated in many IAEA projects or workshops on decommissioning, such as R2D2P, DESA/FASA, IDN, DACCORD and regional TC project RER3009. As a result, personnel have been well trained. Based on the activities mentioned above, preliminary HWRR decommissioning plan has been studied and developed. Proposal of HWRR decommissioning has been developed and submitted to the authority for review and approval. In summary, a number of activities have been conducted during the transition period and HWRR decommissioning is well prepared.

Speaker: Mr YIDONG ZHOU (CHINA INSTITUTE OF ATOMIC ENERGY)

09:00 USING ALPHA AND BETA LIQUID SCINTILLATION COUNTING FOR SAMPLE SELECTION IN DECOMMISSIONING PROJECTS

During the decommissioning of a nuclear site the operator must assure the protection of the workers and the environment. It must furthermore identify and classify the various wastes, while optimizing the costs. At all stages of the decommissioning radiological measurements are performed to determine the initial situation, to monitor the demolition and clean-up, and to verify the final situation. To address these various operations, direct measurement methods can be used (onsite gamma or X spectrometry, autoradiography). For clean surfaces, contamination detectors can be used (COMO, LB, etc.). But for soils, especially damp ones, direct alpha or beta measurement can’t be performed, preparation is needed. Radiochemical analysis is crucial for the radiological evaluation of soils contaminated by alpha and beta emitting radionuclides such as plutonium and strontium. These analyses are expensive and time consuming, due to the many chemical preparations steps needed to purify the radionuclide to be measured. A sampling plan is needed to get an accurate characterization of the pollution. In this way, alpha liquid scintillation counting can be a precious tool. This paper describes a study performed to highlight the capacity of the alpha LSC to detect abnormal counting rates in soils. Standard counting rates has been measured in clean soils then compared to soils containing plutonium and/or strontium. This protocol, including short preparation and high efficiency detection reveal itself useful and time-saving. This study was performed in the SMRT mobile laboratories of the expertise platform in Fontenay-aux-Roses.

Speaker: Mr vincent goudeau (CEA)

09:00 → 11:00 Session 4B - 1: Technical and Technological Aspects of Implementing ER Programmes - Parallel Session

The purpose of this session is:
• To review the available technologies for environmental remediation
• To define the existing gaps in knowledge and needed improvements with the aim of facilitating the implementation these activities.

09:05 Using Geostatistics to Improve Understanding of Contaminated Land Legacies – A Case Study at Sellafield

Sellafield is the UK facility for Nuclear Fuel Reprocessing and Waste Management. It is a compact coastal site with an area of around 3 km2. It is currently operational and is expected to remain licensed until 2120. Radioactive material has entered the sub-surface environment during operations following accidental leaks. This material is currently under active risk management prior to a final hazard reduction and site remediation phase. Sellafield Ltd has to understand and control the legacy of ground contamination to ensure protection of the workforce, the public and the environment. The main control exercised over this material is through an extensive monitoring and risk modelling programme. This work generates a quantity of important environmental data gathered at public cost. Features of the data that make its interpretation difficult include: • Very large data sets • Many different parameters • Long time span (several decades) • Variable quality (changes in analytical methods over time) • Spot 3-D data (as opposed to continuous or 2-D plots) To ensure that the best use is being made of appropriate methods for the gathering and understanding of these data, Sellafield Ltd has commissioned geostatistical analysis. Soils and groundwater data are necessarily spatially correlated and require dedicated geostatistics data processing. Different spatial anisotropies are observed in the saturated and non-saturated zones and integrated in the model. Uncertainty quantification of contaminated volume estimates according to several radiological waste thresholds is addressed to improve risk analysis (remediation feasibility, costs, waste management…). Finally, a critical review of the sampling effort identifies under- or over-sampled areas based on the spatial auto-correlation description. Improvements have been achieved in the following areas through this work: • Inventory – better quantification of contaminated land volumes with an understanding of uncertainty • Characterisation – greater resolution of contaminant distribution with an understanding of uncertainty • Modelling - development of predictive transport models reflecting heterogeneous conditions • Management – improved use of data visualisation to support communication, planning & risk management The paper will address: • The background to Sellafield land quality data interpretation challenges • A description of the use of geostatistics to analyse Sellafield datasets • A discussion on how results could be used in the future and on potential applications of geostatistics to other nuclear site challenges

Speakers: Mr Julian Cruickshank (Sellafield Ltd), Yvon Desnoyers (Geovariances)

09:30 Applications of Ecological Engineering Remedies for Uranium Processing Sites, USA

The U.S. Department of Energy (USDOE) is responsible for remediation of environmental contamination and long-term stewardship of sites associated with the legacy of nuclear weapons production during the Cold War in the United States. Protection of human health and the environment will be required for hundreds or even thousands of years at many legacy sites. USDOE continually evaluates and applies advances in science and technology to improve the effectiveness and sustainability of surface and groundwater clean-ups. This paper is a synopsis of ecological engineering applications that USDOE is evaluating to assess the effectiveness of clean-ups at former uranium processing sites in the southwestern United States. Ecological engineering remedies are predicated on the concept that natural ecological processes at legacy sites, once understood, can be beneficially enhanced or manipulated. Advances in tools for characterizing key processes and for monitoring remedy performance are demonstrating potential. We present test cases for four ecological engineering remedies that may be candidates for international applications: hydraulic control of groundwater, in situ plant and microbial remedies for soil and groundwater, land-farm phytoremediation of groundwater, and evapotranspiration covers for tailings disposal cells. Hydraulic Control of Groundwater USDOE is responsible for characterizing and remediating groundwater at several former uranium mill sites. Groundwater contamination at these sites is attributable primarily to large volumes of processing liquids that seeped from tailings impoundments during the years that mills operated. We evaluated evapotranspiration (ET) by native plants to hydraulically control groundwater flow as an alternative to pump-and-treat remedies at three sites in Arizona and New Mexico. We characterized the plant ecology of sites, strategically transplanted native desert phreatophytes, and developed an empirical algorithm that combines satellite imagery and ground data to estimate landscape-scale ET. Results show that by managing livestock grazing and planting native phreatophytes, we can use ET to control upland recharge, enhance groundwater discharge, and thereby sustainably help control groundwater flow and contaminant transport. In Situ Plant and Microbial Remedies for Soil and Groundwater Uranium processing fluids leaching from tailings left residual contamination in soil and groundwater at many legacy sites. We combined phytoremediation and microbial cycles to reduce nitrate and ammonium levels at a former uranium mill site in Arizona. Contaminants were leaching into groundwater from a denuded soil area where a tailings pile had been removed. We planted and deficit-irrigated this source area with two species of native shrubs, and then discontinued irrigation. ET curtailed leaching, and total soil nitrogen levels dropped >80% over 15 years. Nitrogen isotope analyses indicated that the drop was attributable to coupled microbial nitrification and denitrification processes. We also greatly enhanced rates of microbial denitrification in groundwater by injecting ethanol, which also reduced sulfate and uranium levels and led to our current investigation of ethanol injection to enhance biosequestration of uranium in groundwater at the site.1 Land Farm Phytoremediation of Groundwater USDOE evaluated land-farm phytoremediation as a pump-and-treat approach for nitrate, ammonia, and sulfate contamination in groundwater at a former mill site in Arizona. We irrigated a planting of native shrubs with nitrogen-contaminated groundwater pumped from an alluvial aquifer. Plant uptake and microbial denitrification and nitrification cycles kept nitrogen levels from building up in the land-farm soil, plant growth and transpiration limited recharge and leaching of nitrate and ammonia back into the aquifer, sulfate pumped from the plume remained in the soil profile sequestered as calcium sulfate, and the land farm produced a native seed crop that indigenous people could use for rangeland revegetation and mine land reclamation. Evapotranspiration Covers for Tailings Disposal Cells USDOE is evaluating ET covers as an alternative to conventional covers for tailings disposal cells. ET covers consist of thick, fine-textured soil layers that retain precipitation, which is seasonally removed by plants. Capillary barriers composed of coarse-textured sand and gravel placed below this soil “sponge” can enhance soil water storage capacity and limit unsaturated flow. The sustainability of ET covers depends, in part, on the establishment and resilience of a diverse plant community. We used a series of increasingly larger lysimeters to design an ET cover to enhance the performance of a uranium mill tailings disposal cell in Utah. The design used sandy clay loam soil from the site and native shrub-steppe vegetation. Lysimetry offers the only direct means for measuring percolation at field scale and allows comprehensive evaluation of the soil water balance. Percolation, as recorded with a 3-ha drainage lysimeter installed within the cover during construction, was approximately 0.1% of precipitation during 15 years of monitoring. 1 Biosequestration generates reducing conditions in groundwater by stimulating the growth of microbial populations through injection of electron donor compounds into the subsurface.

Speaker: Dr William Waugh (US Department of Energy)

09:55 ASSESSMENT OF SPECIATION AND MOBILITY OF URANIUM IN ABANDONED TAILING SITE IN UKRAINE

For decision making in choosing remediation strategies on Uranium legacy sites it is vitally important to know whether the radioactive properties of the tailing material are stabile in changing environment. This report presents the results of investigation carried out on Centralny Yar (CY) tailing, one of the oldest tailings of the Pridneprovsky Chemical Plant (PChP), Dniprodzergynsk, Ukraine, formed during 1950s. After filling the tailing reservoir located at the Dnieper River terrace ravine with radioactive waste of Uranium production, it was covered by a mixture of construction debris and soils, cultivated and planted with trees and bushes. CY tailing (area 2.4ha) contains 0.22 million tons of waste material with estimated gross radioactivity as 104 TBq. Distinctive feature of this tailing is strongly acidic reaction (pH 2.5 - 4) within the tailing body due to filling with primarily processed uranium extraction residue reportedly without applying chemical neutralization. There was no special engineering protective cover designed for both lower and upper interface of the tailing body. In order to understand how changing environment could affect Uranium migration potential within the tailing body a method of determination Uranium chemical speciation in stored material have been deployed. In frame of 2012 Ukrainian National Program of PChP rehabilitation geological boreholes were drilled on CY for site characterization. Samples of core material collected during drilling were subjected for further radionuclide content analysis and set of characteristic over the vertical profile samples were chosen for extended assessment. This comprised of determination chemical speciation of radionuclides in the tailings material using modified method of sequential extractions known as BCR (developed after EC Bureau of Certified Reference Material). The 4-step experimental protocol is designed to evaluate water-soluble, acid soluble, reducible and oxidizable chemical speciation of radionuclides. Measurements of gamma-emitting radionuclides in solid samples and extraction aliquots were made using low-background HPGe detector GMX40C, ORTEC. For validation the results some representative samples have been analyzed for Uranium content using radiochemical separation and alpha-spectrometry. Vertical profile of the radionuclides distribution in boreholes evidence existence of clearly defined stratification within the tailing body due to differences of migration properties of chemical elements in certain geochemical conditions. It is primarily concerned with most mobile Uranium (measured as U-238) and least mobile Radium (measured as Ra-226). It was found that the ratio between the two isotopes in the upper part is 5, reaching 25-30 in the middle, and less than 1 at the bottom section of the tailing. Furthermore, it was found that the front of Uranium distribution is shifted down by ~2m against Radium and its maxima correspond to the most water-saturated layer. Taking into account specific geochemical conditions developed within the tailing body, it is reasonable to combine water and acid soluble speciation of Uranium, derived by BCR extraction protocol, into a single group (readily exchangeable form). This fraction of radioactivity could be effectively washed out from the tailings body to the local aquifer by infiltration of rainwater. With next steps of the sequential extraction protocol less soluble form of uranium could be estimated. Reducible fraction represents Uranium bound to Fe, Mn oxides and hydroxides complexes and oxidizable - in sulfides and sparingly soluble organic complexes. Residual after extraction contains Uranium in practically insoluble form. It is clearly seen that distribution of different forms of Uranium in tailing material follow general pattern when amount of exchangeable form increases towards the bottom of the tailings. Taken that in strongly acidic conditions the main constituent in groundwater composition within and under the tailings is sulphate-ion (up to 2.4 g/l of SO42-) one can assume that Uranium predominantly migrates in form of Sulphate. Centralny Yar Tailing is one of the topic object of radioecological concern at former PChP. Strongly acidic reaction of the media within its body promote Uranium migration outside the core into the groundwater. According to the experimental results Uranium migrates in tailing mainly in form of Sulphate. Application of sequential extraction protocol BCR revealed that the soluble form of Uranium exceed 10% of the total content in present conditions, increasing to 20-40% in ultimate geochemical composition of harsh ion exchange.

Speaker: Mr Kyrylo Korychenskyi (Ukrainian Hydrometeorological Institute (UHMI), Department of environment radiation monitoring)

10:20 Vehicle-Mounted Gamma Spectrometer for Radiological Land Surface Surveys

Abstract: A vehicle-mounted gamma spectrometer, Mobile Gamma Spectrometer (MGS), that is a practical solution for surface soil monitoring in rough, desert terrain is described. Surface soil scanning is applicable to land characterization, monitoring and clearance, all potential end-points of environmental remediation. The MGS is constructed, almost entirely, from proven commercially-available components and was successfully used to survey approximately 81 ha of desert for small areas of elevated radioactivity (hotspots). The results were compared against those from background areas and against a model hotspot that would contribute 20 µSv/y under an industrial scenario. Data processing was carried out with Mathematica 10, which includes strong list-processing and graphics capabilities, as well as a functional language for general mathematics. Two technicians can scan about 0.4 ha/hr. The system, with its light weight and wide tires, does minimal damage to the desert environment. (Note: Synopsis is attached.)

Speaker: Mr William Millsap (Dade Moeller & Associates (US Dept. Energy))

10:25 Contaminated Land Remediation on decommissioned nuclear facilities : an optimised approach

The effluent treatment facility on the decommissioned nuclear power plant of Brennilis, France, is currently being dismantled. Historically, various incidents involving effluent drum spills caused radiological contamination in the building platform and the underlying soil. The French Nuclear Safety Authority’s doctrine for radioactively contaminated land is a reference approach which involves complete clean-up, removing any trace of artificial radioactivity in the ground. If technical difficulties are encountered or the quantity of radioactive waste produced is too voluminous, an optimised clean-up can be implemented. In our case, the reference approach being disproportionate to the risks, an optimised approach is proposed. EDF first chose to demonstrate compliance between the land and all foreseeable uses. It then involved optimising earth excavation to reduce the source term as far as reasonably achievable, taking into account technical difficulties, quantities of low level waste produced, sustainable management and workers’ safety.

Speaker: Mrs Emilie SAUER (EDF DIPDE)

10:30 Experience of FSUE “RADON” in rehabilitation of the legacy sites and radwaste

Since 1961 FSUE RADON makes its activity in the field of Radwaste management, radiation safety and rehabilitation of environment. The first transportation of Radwaste to the site was implemented on January, 27, 1961. Now FSUE RADON serves the central part of Russian Federation and nearest regions. The fields of RADON's activity are the following:  Radiation-Ecological Monitoring,  Decommissioning Nuclear Facilities and Rehabilitation Of Areas,  Radwaste management: Collection, Transportation, Processing, Conditioning and Storage of Radioactive Waste,  Engineering, Scientific and Technological Support. The company carries out the complex managing radioactive waste, radiation examination and radiation-ecological monitoring of environmental objects. The main activities in the decommissioning are:  Unload and removal of spent nuclear fuel and nuclear materials from the site (transfer of nuclear facility into a safe condition);  Removing the working environments;  Decontamination of equipment and facilities;  Dismantling, demolition of constructions and buildings;  Treatment of the resulting radioactive waste, their containerization and delivery for storage and disposal;  Remediation of contaminated sites. Decontamination and rehabilitation of facilities and territories  Development of a special part of the project documentation for the decontamination and rehabilitation of contaminated sites and areas  Collecting, sorting and preparation for transportation of radioactive waste  Rehabilitation facilities and areas for further use without restriction  Control radiation inspection facilities and areas after the completion of decontamination and rehabilitation The following legacy sites were decommissioned and cleaned up:  Rehabilitation of the Applied Chemistry Institute, St. Petersburg: Dismantling of the concrete canyon of the former storage of liquid radioactive waste, packaging waste and their transportation to the interim storage site, containerization and removal of 400 m3 radioactive waste  Dismantling the plant and rehabilitation of area of "Podolsk factory of non-ferrous metals" Moscow region.  Rehabilitation of "Globus 1" (Ivanovo region): Excavation and sorting of 2415 m3 contaminated soil, containerization and removal of 800 m3 radioactive waste.  Rehabilitation of Areas of Prospecting and Experimental Works (Kurgan Region): Examination of technological bores of reservoir for situ leaching of uranium and adjacent territories, land clearing and liquidation plugging of decommissioned 72 technological wells, Remediation of an area of 3000 m2: Filling of contaminated areas of a private ground; The reference radiation survey of the rehabilitated area; Planting herbs at the rehabilitated areas.  Rehabilitation of the Experimental Field in the Vladimir Region

Speaker: Mr DENIS FEDOROV (FSUE "RADON")

10:35 MULTI-CRITERIA OPTIMIZATION TOOL FOR THE REMEDIATION OF URBAN AREAS AFTER A NUCLEAR ACCIDENT

This paper describes the development of a multicriteria tool to support decision making process after a nuclear accident that leads to the contamination of urban areas. The methodology adopted is discussed and the result is a computer program that assess doses to residents and to remediation workers as a function of time, the effect of remediation procedures on the doses to public and estimates wastes generated for pre-defined scenarios. Some preliminary results are presented based on remediation procedures used after the Goiania accident.

Speaker: Mr Christiano De Luca (UFRJ)

10:40 Geostatistics for radiological characterization: Sampling optimisation, data interpretation and risk analysis

Need for proper characterisation Dismantling and decommissioning of nuclear facilities or remediation of contaminated sites are industrial projects with huge challenges. Precise knowledge of the contamination state is required. Radiological evaluations have multiple objectives to be considered: determination of average activity levels, to allow the categorization of surfaces or volumes (sorted into different radioactive waste categories); location of hot spots (small areas with significant activity levels); and estimation of the source term (total activity) contained in soils or building structures. In addition there are radiation protection and other logistics considerations. Estimates are essential for the proper management of these projects. Currently, characterization remains relatively empirical. Accumulated approximations often have serious consequences that threaten the project’s successful completion, for example through over-categorization or unexpected contamination. Radioactive contamination is generally complex and involves numerous parameters: radiological fingerprint, transfer path, type of contaminated materials, presence of different matrices (soils, concrete), and so on. Numerical modelling often turns out to be very difficult. The characterization phase should be efficient and the sampling strategy has to be rational. However, investigations also represent capital expenditure; the cost of radiation protection constraints and laboratory analysis can represent a large amount of money, depending on the radionuclide. Therefore the entire sampling strategy should be optimized to reduce useless samples and unnecessary measures. Geostatistics methodology The geostatistical approach, which provides consistent estimates and reliable maps, is an appropriate solution for data analysis. Geostatistics aims to describe structured phenomena in space, possibly in time, and to quantify global or local estimation uncertainties. Estimates are calculated from a partial sampling and result in different representations of the contamination, including interpolation mapping (‘kriging’). But the added value of geostatistics goes beyond this. Its benefit is its ability to quantify estimation uncertainty and provide risk analysis for decision making. More advanced and sophisticated geostatistical methods, such as conditional expectation or geostatistical simulations, can be used to quantify risk of exceeding the threshold, for instance. These estimates are powerful decision-making aids when classifying surfaces and volumes before decontamination starts (based on different thresholds as well as considering the remediation support impact). Finally, multivariate geostatistics allows different kinds of information to be combined to improve estimates, using the spatial correlations between variables. Physical and historical data and non-destructive measurement results (for example dose rate or in situ gamma spectrometry) are integrated to improve understanding and prediction of the main variable (results of laboratory analysis, for example) while reducing the estimation uncertainty. Data consolidation and sampling optimisation To use geostatistics, datasets must be consistent for correct data processing: the same sampling protocol must be used, measurement or analysis must be performed in a short period if the decay rate is sensitive; data of the same type must be expressed in the same unit, and so on. This may seem obvious, but a lot of time can be lost in correcting errors and ensuring that the data are really consistent: coordinates, dates, units, physical and radiological heterogeneities, odd correlations… In addition, the spatial structure of radioactive contamination makes the optimization of sampling (number and position of data points) particularly important. Geostatistics methodology can help determine the initial mesh size and reduce estimation uncertainties. Application cases and lessons learned Geostatistics can be applied to a range of radiological characterization. The only limitation is the quality of the input data, since they are fundamental to describing spatial structure of the phenomenon. Geostatistics maps can cover very small areas (a few m² or a few m³) or very large sites (at a country scale) as presented in several application cases: characterization of a legacy site, post-accidental mapping, building structure categorization… To conclude, geostatistics is a very powerful tool to analyze, consolidate and give value to collected pieces of information. The exploratory data analysis, in combination with the spatial structure interpretation (variogram) is probably the most interesting part of the characterization. It results in mapping outputs, risk analysis and sampling optimization.

Speaker: Dr Yvon Desnoyers (Geovariances)

10:45 MOBILE LABORATORIES: AN INNOVATIVE AND EFFICIENT SOLUTION FOR RADIOLOGICAL CHARACTERIZATION OF SITES UNDER OR AFTER DECOMMISSIONING AND ENVIRONMENTAL REMEDIATION PROGRAMMES.

During the operation and the decommissioning of a nuclear site the operator must assure the protection of the workers and the environment. It must furthermore identify and classify the various wastes, while optimizing the costs. At all stages of the decommissioning radiological measurements are performed to determine the initial situation, to monitor the demolition and clean-up, and to verify the final situation. Radiochemical analysis is crucial for the radiological evaluation process to optimize the clean-up operations and to the respect limits defined with the authorities. Even though these types of analysis are omnipresent in activities such as the exploitation, the monitoring, and the cleaning up of nuclear plants, some nuclear sites do not have their own radiochemical analysis laboratory. Mobile facilities can overcome this lack when nuclear facilities are dismantled, when contaminated sites are cleaned-up, or in a post-accident situation. The current operations for the characterization of radiological soils of CEA nuclear facilities, lead to a large increase of radiochemical analysis. To manage this high throughput of samples in a timely manner, the CEA has developed a new mobile laboratory for the clean-up of its soils, called SMaRT (Shelter for Monitoring and nucleAR chemisTry). This laboratory is dedicated to the preparation and the radiochemical analysis (alpha, beta, and gamma) of potentially contaminated samples. In this framework, CEA and Eichrom laboratories has signed a partnership agreement to extend the analytical capacities and bring on site optimized and validated methods for different problematic. Gamma-emitting radionuclides can usually be measured in situ as little or no sample preparation is required. Alpha and beta-emitting radionuclides are a different matter. Analytical chemistry laboratory facilities are required. Mobile and transportable laboratories equipped with the necessary tools can provide all that is needed. The main advantage of a mobile laboratory is its portability; the shelter can be placed in the vicinity of nuclear facilities under decommissioning, or of contaminated sites with infrastructures unsuitable for the reception and treatment of radioactive samples. Radiological analysis can then be performed without the disadvantages of radioactive material transport. This paper describes how this solution allows a fast response and control of costs, with a high analytical capacity.

Speaker: Mr PATRICE LETESSIER (EICHROM LABORATORIES)

09:00 → 18:00 Session 4B - Poster

09:00 DECONTAMINATION METHOD AGAINST 137CS IN ORCHARD SOIL BY COMBINING UNDERGROWTH AND LAYING MATERIALS

Effects of decontamination method against 137Cs in orchard soil by combining undergrowth and laying materials covering over soil surface were evaluated. Sheets containing zeolite or Prussian blue, root wrapping nets , revegetation nets , undergrowth, such as shepherd’s-purse, deadnettle, bittercress and others, and grosses for Kentucky Bull Glass and white Clover, which were sowed seeds mixed 40 g/m2 and 30 g/m2 on the sheets on May in 2012 or March in 2014, were selected for investigation. Soil surface was covered with laying materials which is 60 or 190 cm in width. After nourishing the undergrowth in natural or grasses which were sowed seeds, laying materials were wound up with foliage and root of undergrowth or grass and rhizosphere soil for stripping the topsoil. Removal rate of 137Cs(%, RR), which represented the following formula: RR=100a/(a+A), a is 137Cs content (kBq/m2) in removed soil, A is 137Cs content in 3 cm soil layer under the topsoil stripped, was applied for evaluating the effect to decontamination of 137Cs in soil. Revegetation nets with gross showed significantly higher RR, resulting in 16.7% by sowing on May and 35.7 to 53.3% (removed 45.9 to 60.6% 137Cs compared with control) by sowing on March. RR of Kentucky Bull Glass with root mat developed sufficiently attained to 94.1%, indicating that decontaminating effect was due to development of the root mat layer. On the other hand, increase of RR was expected by increasing density of mesh of revegetation nets. A proportional relation expressed y-6.07x in dry weight between removal weight of soil (x) and RR (y) was recognized significantly.

Speaker: Mr Mamoru Sato (Faculty of Symbiotic Systems Science, Fukushima University)

09:00 Environmental Remediation Programm

There were in Azerbaijan Iodine Plan Production. After the farmer of Soviet Union these plant were not continue work. The the some territory has been contaminated by charcoal. After the degree of Azerbaijan President Ilham Aliyev these contaminated territory began to clean from charcoal. And all these work provide by the Ministry of Emergency Situations of Azerbaijan Republic. All the contaminated soil were transported by vehicle to the site of Specialized Establishment "Isotope" of the Ministry of Emergency Situations of Azerbaijan Republic.

Speaker: Mr Aslanov Fikrat (Azerbaijany)

09:00 Environmental remediation programmes

After the firmer of SSRI in Azerbaijan Republic have two iodine production plant stop you work. And, some territories of Absheron penunsila has been contaminated by charcoal. After the degree of Azerbaijan President of Ilham Aliyev's the contaminated territory began to clean by the Ministry of Emergency Sitautions of Azerbaijan Republic with IAEA. IAEA's international experts give us more recimendation about clean contaminated territory and help us in this way. Now this territory has been cleaned and our goverment create a new park for the local people and the people and children can rest in these territories. Then our government decide to clran other territory which located 200 km far from the capital of Azerbaijan Republic of Baku Neftchala Iodone Brome Plant. This territory also cleaned by the recomendations and advices of IAEA's experts.And this territory also has been cleaned. All cintaminated soil transported to the area of the Specialized Establishment "Isotope" of the Ministry of Emergency Situations of Azerbaijan Republic and prepare the well for them and long term storaged.

Speakers: Mr Aslanov Fikrat (AZERBAIJAN), Mr Ceyhun Eyvazov (Azerbaijan)

09:00 EVIDENCE AND DOCUMENTATION OF REMEDIATION SUCCESS ACHIEVED AT FORMER URANIUM PRODUCTION SITES IN GERMANY

Abstract: At the advanced state of remediation of complex legacies left behind by uranium mining and milling, the need arises to provide and document evidence of re-medial success. In the case of large amounts of radioactive materials remaining near the sur-face evidence can only be provided in the framework of a prolonged monitoring process. Objective documentation of the remedial success as part of the knowledge management is a pre-condition for the release of rehabilitated objects from regulatory control. Such documentation also provides the basis for implementing long-term tasks on rehabilitated objects including institutional control. This paper illustrates by case examples the way in which evidence of the remedial success is provided and the required data and information on remediation are made available to various users. 1. INTRODUCTION For 25 years now, the federally-owned Wismut GmbH has been remediating the legacies left behind by former uranium ore mining and processing operations in the East of Germany. In that area, the former Soviet-German stock company SDAG Wismut had produced a total of 216.000 tonnes of uranium during a period of more than forty years whereby it had evolved into the world’s fourth largest uranium producer at that time. As a result of mining low-grade ores, more than 800 million tonnes of low-level radioactive materials were deposited in densely populated areas, in some cases in the immediate neighbourhood of residential areas. This created a situation which required in situ remediation associated with residual radioac-tive materials left after clean-up. Remedial progress achieved so far makes it imperative to substantiate and document the remedial success. 2. CRITERIA OF THE REMEDIAL SUCCESS Criteria whereby a remedial action can be considered as successful are both site-specific and object-specific. With regard to achieving environmentally relevant remedial targets at objects with residual contamination left at the site, the efficiency of technical barriers is of particular importance. Especially in the case of covers it revealed that even a long time after their placement they were still subject to alterations, such as in terms of water balance or gas per-meability. As a consequence thereof, contaminant releases via seepage water and radon exha-lation remain in the longer term susceptible to alterations. Therefore, Wismut has identified the criterion for remedial success that the cover performance must have achieved an stable condition, the intended degree of performance is maintained and the risk of negative trend reversal can be precluded. 3. APPROACH TO PROVIDING EVIDENCE OF REMEDIAL SUCCESS. In an effort to provide evidence of remedial success WISMUT first and foremost applies methods of source monitoring (e.g. measurements of contaminant release via seepage, radon exhalation measurements). In addition, methods of source as well as person related environ-mental monitoring are used. These methods allow establishing the residual and overlapping impact of all kinds of remediated objects at relevant exposure sites. Establishment of cover performance oftentimes requires complex investigations of the cover water balance and gas permeability. For a long period of time after cover placement was completed, vegetation suc-cession and various forms of bio-intrusion will continue to produce alterations. To some ex-tent highly sophisticated investigation techniques were used in order to understand the proc-esses going on in covers and to provide evidence of the remedial success [1]. 4. CASE EXAMPLES Evidence of the remedial success is demonstrated by the case examples „Evolution of the radon situation“ and „Evolution of contaminant release via seepage water“, exemplified by reclaimed waste rock piles at the Schlema-Alberoda site. Complex investigation methods included among others: - Lysimeter investigations to quantify water balance of the cover; - Tracer gas applications to analyse airflow within the system waste rock pile – cover as well as infrared investigations to identify air leaks; - Time- and site-resolved measurements of gas permeability and radon exhalation rate. 5. DOCUMENTATION OF THE REMEDIAL SUCCESS WISMUT has devised the technical data base system AL.VIS(W) [2]. The environmental data base is also part of the system. Remediation documentations contain all information required for tracking reme-dial steps and the state of remediation of individual objects. Data and information which con-tinue to be accumulated from post-remedial activities are periodically integrated. The web-based platform AL.VIS(W) enables Wismut as well as authorities and other authorised users to access data and information. REFERENCES [1] SCHMIDT,P., REGNER,J., Improvement of the Radon Situation at Former Uranium Mining Sites in East Germany, 6th Conference on High Levels of Natural Radiation, Osaka, 2004, Internat. Congress Series Vol. 1276 (2005) 238-240. [2] KREYSZIG,E., GÖTZE,J., The Wismut Technical Information System, International Conference WISSYM_2015, Bad Schlema, Reclaimed Mining Sites between Post-Remedial Care and Reuse, ISBN: 978-3-00-049882-4 (2015) 345-347.

Speaker: Dr Peter Schmidt (Wismut GmbH)

09:00 EXPERIENCE SEARCHING OF CONTAMINATED PAST SOVIET UNION MILITARY SITES AND REMEDIATION OF IT IN LITHUANIA

In Lithuania some territories and sites in past were used by soviet army. Sites where industrial, medical or other installations that used sources of ionizing radiation took place in Soviet times as well. Those sites are of interest for search of orphan sources or radioactive contamination. At some of the sites search was performed before the soviet army leaved the sites. Mostly of the soviet type industrial, medical or other purpose installations were decommissioned, but problem exist due to incomplete information about the sources of ionizing radiation. For that reasons the companies of search of orphan sources and contaminates sites started in 2009. Periodical programs dedicated for the search of sites are approved by Director of Radiation Protection Centre. First step was analysis of information available about the sources of ionizing radiation at “old’ industrial, medical or other installations those were operating in soviet time. That information was available at Radiation Protection Centre. Information about the soviet army sites and possible usage of the sources on those sites was obtain from data available at data reports of State Geology Service. During 2009 – 2015 year search was performed at 15 territories and sites in totally around 100 ha. In 2013 year in one of the military site during the search radioactive contamination of ground was found. Preliminary results showed that the site is contaminated with 226Ra. In the most radioactive spots soil sampling was performed and 12±1 Bq/g concentration of 226Ra was measured by gamma spectrometry. Exemption level of 226Ra in Lithuania is 10 Bq/g. By the requirements of legal acts of Lithuania the site should be investigate and assessed. The territory belongs to one of the Lithuanian municipality, and according the legal acts the local municipality shall organize the investigation of contaminated area and if needed the municipality shall organize the remediation as well. Municipality contracted specialists for investigation of contamination at territory and performance of assessment of needs of remediation. The site investigation showed that maximum value of equivalent gamma dose rate at 1 m from the surface of ground was 0.5 µSv/h, at 5 cm from the surface 3 µSv/h. The maximum concentration of 226Ra was 5 Bq/g at 30 cm deep. Based on the investigation RPC recommended to perform on-site remediation of the area. In 2014 year municipality remediated the area. Specialists of Radiation Protection Centre gave advice and suggestions for radiation protection measures during remediation process and made final radiological measurements after remedial measures were performed.

Speaker: Mr Remigijus Kievinas (Radiation Protection Center)

09:00 IODINE BIOTRANSFORMATION: A HANFORD PERSPECTIVE

Radioiodine (129I ), a waste product from nuclear fission, is of environmental concern due to its long half-life (~16 million years), mobility, and hazardous potential to humans through bioaccumulation in the thyroid gland. The Hanford Site in Richland, Washington, contains two separate iodine-129 contaminant plumes over 1,500 acres with concentrations of ~3.5 pCi/L in groundwater samples, exceeding the federal drinking water standard of <1pCi/L. Although, iodide (I-) is thermodynamically favored, based on current pH and Eh ranges measured in the groundwater, the dominant species is iodate (IO3-) (70.6%) while organo-iodine (25.8%) and iodide (3.6%) were found in far lower quantities. Enrichments of Lower Ringgold sediment from the 200 West Hanford site allowed the isolation of microbial species capable of iodine biotransformation. Through a series of batch studies and spectrophotometric assays, isolates were found to couple nitrate (NO3) reduction with iodate (IO3), where iodate reduction was not observed in the absence of nitrate. Additionally, isolates able to oxidize iodide were also identified. Currently, analytical technics are being developed to further understand the kinetics and enzymatic activity for both of these redox reactions. Ongoing research involves these isolates and their influence on iodine speciation in the presence of organics such as humic acid or lignocellulose. 1. INTRODUCTION From the 1940s through the early 1990s, liquid wastes from materials used and produced at the Hanford Site were disposed to the ground through cribs, ditches, ponds, and trenches. Waste containing carbon tetrachloride, uranium, nitrate, chromium (total and hexavalent), I-129, Tc-99, and tritium infiltrated the subsurface contaminating soil and groundwater [1]. Iodine-129 is one of the primary risk drivers for the site. Hydraulic containment is the currently selected remedy for I-129 in the groundwater, and there is currently no remedy selected for controlling migration of I-129 from the vadose zone to the groundwater. A number of biogeochemical processes have been shown to cycle iodine between different aqueous, as well as volatile and solid phase species (Figure 1) [2]. The effect to microbes on iodine transformation by Hanford isolates will be discussed. 2. METHODS Batch microcosm studies were performed to look at the ability of bacteria isolated from the Hanford subsurface to transform different iodine species when supplied with an external carbon source. In separate studies, bacteria were supplied with lactate, nitrate and iodate, and bacterial growth and iodate disappearance were monitored with time. In a similar manner, bacterial isolates were supplied with either an organic acid or sugar with iodide and iodine concentration was monitored with time. All analytes were measured using colorimetric methods. 3. RESULTS Molecular characterization of the microbial community in iodine-impacted groundwater showed a number of bacterial families with genera that have previously demonstrated iodine biotransformation capabilities. Results from sediment traps incubated in locations in the plume with background, low, and high levels of 129I show the presence of families that include Pseudomonads and Actinobacteria, taxa that have shown the ability to both oxidize I- and reduce IO3-. Sediments from traps incubated in iodine-contaminated groundwater at the Hanford Site have yielded a number of bacterial isolates that can oxidize or reduce different iodine species. Since the dominant iodine species in 200-UP-1 groundwater has been shown to be IO3-, experiments were performed to determine the ability of various Hanford isolates to reduce IO3- in the presence of nitrate, a common co-contaminant in the 200-UP-1 groundwater. One isolate designated AD35, which is most closely related to Agrobacterium tumefaciens, has been shown to reduce IO3- to I- in the presence of nitrate. Iodate reduction occurred under both anaerobic and microaerobic conditions. When the culture was spiked with nitrate, IO3- concentrations continued to decrease in the culture medium. Microbial oxidation of I- to I2 and IO3- is another important process that could affect iodine speciation in the Hanford groundwater. While I- concentrations in Hanford groundwater are below 5% of the total iodine, the potential for biological I- oxidation will provide insight into current speciation, and prevalence of IO3-, in the 200 West Area groundwater. A number of bacterial isolates from the Hanford groundwater have been shown to oxidize to I- to I2. 4. CONCLUSIONS Bacteria isolated from Hanford sediments were shown to be capable of iodine cycling in the subsurface. These results are important because: 1) bacteria capable of iodide oxidation show that microbes may be responsible for the current iodine speciation in the Hanford subsurface; 2) bacteria can change the speciation of iodine; thereby, affecting fate and transport of iodine in the subsurface; and 3) bacteria could be used as part of remedial strategies for future cleanup of iodine in the Hanford vadose zone and underlying groundwater.

Speaker: Dr Michelle Hope Lee (Pacific Northwest National Laboratory)

09:00 LORADO URANIUM MINE ENVIRONMENTAL REMEDIATION – NORTHERN SASKATCHEWAN

LORADO URANIUM MINE ENVIRONMENTAL REMEDIATION – NORTHERN SASKATCHEWAN SUBMITTED TO THE 2016 IAEA INTERNATIONAL CONFERENCE ON ADVANCING THE GLOBAL IMPLEMENTATION OF DECOMMISSIONING AND ENVIRONMENTAL REMEDIATION PROGRAMMES WILSON, Ian; ALLEN, Dianne E.; SCHRAMM, Laurier L.; and MULDOON, Joseph Saskatchewan Research Council, Saskatoon, SK, Canada E-mail address: ian.wilson@src.sk.ca 1. INTRODUCTION The abandoned Lorado mine and mill sites are located in northwestern Canada, north of Lake Athabasca, on the western shore of Nero Lake, and 8 km south of Uranium City (Figure 1). During its operating lifetime the mill processed 305,000 to 550,000 t of low-grade ore to produce about 690 t of U3O8 product, and 190,000 to 344,000 m3 of tailings. The mill tailings and acidic waste were initially deposited in a nearby depression, but eventually extended beyond that to cover about 14 ha, and then overflow into nearby Nero Lake. Following closure in 1961 the mine and mill were dismantled over a period of years, and buried, but no remediation was undertaken of the exposed tailings or of Nero Lake. In 2007 the Saskatchewan government retained the Saskatchewan Research Council (SRC) to develop a remediation plan, undertake an environmental assessment of the proposed project, and manage remediation activities at the Lorado site. The resulting environmental impact statement describes a risk reduction plan designed to provide net benefit to human health, wildlife populations and aquatic life. The principal risks to humans, wildlife, and aquatic life were determined to arise from the exposed tailings and contaminant flows from Nero Lake to another nearby lake, Beaverlodge Lake. Accordingly, the principal reclamation activities were to: cover the surface tailings in-place and batch-treat Nero Lake in situ. The reclamation work has been conducted with approvals, permits and licenses from the Saskatchewan Ministry of Environment and the Canadian Nuclear Safety Commission. 2. SURFACE TAILINGS COVER The objectives of the cover system are to protect against gamma radiation and radon exposure, prevent formation of efflorescent salts on the surface, reduce tailings acid generation in Nero Lake, and block uptake of tailings pore fluid into vegetation. It also had to cover soils and vegetation adjacent to the tailings that may have been contaminated. For exposed tailings an engineered soil cover system was designed that incorporates a “capillary break,” comprising a 1 m thick layer of specifically-sized sand under a 0.25 m layer of finer soil (“till”). The sand layer is to prevent the upward movement of water, and the till layer is to protect the sand cover and allow revegetation. On the submerged tailings within Nero Lake, only a sand cover was required. Overall, the covers used about 250,500 m3 of sand and about 93,400 m3 of till. The cover system was also designed to ensure that it is free draining and contoured to shed water towards one of two engineered drainage ditches, or directly to Nero Lake. The cover system will ultimately be stabilized and revegetated. 3. IN SITU NERO LAKE TREATMENT Nero Lake is about 2.2 km long by about 1 km wide, with a total water volume of about 11 million m3. It is estimated that as much as 167,000 m3 of tailings and acidic waste overflowed into Nero Lake, and that as much as 40% of the bottom of Nero Lake was covered with tailings. Since abandonment, additional contaminants have flowed into Nero Lake. The cumulative effects of this contamination included the lake pH dropping to 4.0, reductions in alkalinity and bicarbonate concentration, and elevated concentrations of dissolved heavy metals such that the lake was no longer able to sustain a fish habitat. A treatment process for the lake was developed, pilot tested, and implemented in 2013-2014. This involved mildly overdosing the lake with about 400 tonnes of lime (CaO). The treatment was successful, and brought the pH of the lake to about 7.5. 4. CONCLUSIONS Concerns that the Lorado mill site has been affecting human health and the environment have been dealt with by completing a remediation project to reduce such impacts from this site in future. An effective remediation plan was developed and implemented, which is based on proven scientific methods, risk assessment, and health, safety and environmental best practices. Monitoring will continue to confirm predictions made during the environmental assessment, and to assist with on-going planning and controls put in place to manage the site.

Speaker: Mr Ian Wilson (Saskatchewan Research Council)

09:00 NEW TECHNOLOGIES FOR REMEDIATION OF RADIONUCLIDE CONTAMINATED ENVIRONMENT

Contamination of the environment in the Republic of Moldova may be caused as a result of the collection, processing, storage and disposal of wastes associated with the use of radioactive isotopes for industrial, research and medical applications. As a result of such contamination, in the Republic of Moldova have been initiated projects to assess and remediate radioactively-contaminated sites. The Chernobyl nuclear accident also affected the territory of the Republic of Moldova. In the early months after the accident, the levels of radioactivity of agricultural plants and plant-consuming animals were dominated by surface deposits of radionuclides. The deposition of radioiodine caused the most immediate concern. Radioisotopes of caesium (137Cs and 134Cs) were the nuclides which led to the largest problems and even after decay of 134Cs the levels of longer lived 137Cs in agricultural products from highly affected areas still may require environmental remediation. Many of the currently available methods for cleanup of the residual contamination are very expensive or ineffective in given situations. The most obvious intervention is the simple removal of contaminated material. Biological processes of in situ treatment technologies have an important economic role [2, 3, 4]. Over 97% and about 85% removal of137Cs and 60Co, respectively, from the pool water were reported, at a flow residence time of 7 days in Korea [6]. It is supposed that microorganisms degrade radioactive substances in a relatively short time (about 15 hours) [5]. The aim of the paper was to develop modern biotechnology for environmental remediation of contaminated with radioactive compounds. Research objectives: monitoring the concentration of natural radionuclides in soils from main geographical areas of the country, including the radioactive waste storage facility; isolating fungi from investigated soils in laboratory conditions; testing of the fungi pathogenicity and selecting nonpathogenic strains; testing in vitro influence of fungi Mucur vulgaris, Penicillium viride and Aspergillus niger on solubilisation of cobalt phosphate compound (CP). Material and methods. Testing of a set (n=46) of fungal strains according to their capacity to dissolve radioactive compounds was conducted. Qualitative assessment of potential biosynthesis able to solubilise CP by strains of fungi was performed by determining the development of fungal colony diameter and the diameter of expansion solubilising area over 24, 48, 72, 94 etc. hours of cultivation on the wort-agar medium (control), and supplemented with the CP in a concentration of 0.5...1.5%. Diameter of fungus colony and diameter of solubilising area around the colony was measured [1] (fig.1). Fig. 1. Penicillium viride colony and cobalt phosphate solubilising area development. Results: It has been demonstrated that all species of studied fungi dissolved FC but their capacity were different. So, in variant with 0.5% concentration of CF, solubilising diameter area under the action of Penicillium viride strain 2 was 90.0±0 mm, with 10.0 mm being greater than if the action of Aspergillus niger strain, and consisted 112.5% compared to the analogue 1. At the same time, in the second variant with the 1% concentration of CP, solubilising diameter area of CF under the action of Penicillium viride strain 2 was 90.0±0 mm, 26.25 mm being higher than if action Aspergillus niger strain. So efficiency consisted 204.5% compared to analogue. Efficiency versus analogue 2 – Mucor vulgaris X varied between 116,1–141,2%. Interesting results have been obtained studding another radionuclide compounds and microbes, which will be reported in poster presentation. Conclusion: The application of bioremediation as a biotechnological process involving microorganisms, especially fungi has great value, because of its increasing potential of solving the dangers of many radionuclide pollutants through biodegradation. Bioremediation using biodegradation represents a high impact strategy, but still a low cost way tool of removing pollutants, hence a very viable process to be applied. The principles of bioremediation are based on natural attenuation/bio augmentation/bio stimulation. Bibliography 1. COREŢCHI L. et all. Fungi strain Mucor vulgaris X for solubilisation of insoluble compounds of cobalt. Patent nr 3212. (in rom.). 2. COREŢCHI, L.; PLĂVAN, I. Interaction of microorganism with radionuclide. Public health, Economy & Management in Medicine, 4/55, 2014, p. 4-14. ISSN: 1729-8687. ( in rom.). 3. DHAN P. et al. Bioremediation: a genuine technology to remediate radionuclides from the environment. Microb. Biotech., 2013, vol. 6, p. 349-360. 4. IAEA, Predisposal Management of Organic Radioactive waste. Technical Rep.Series No.427, IAEA, Vienna, 2004. 5. PANAK P. J. et al. X-ray absorption fine structure spectroscopy of plutonium complexes with Bacillus sphaericus. In: Radiochimica Acta, 2002, vol. 90, nr. 6, p. 315-321. 6. PATERSON-BEEDLE M. et all. Utilisation of a hydrogen uranyl phosphate-based ion exchanger supported on a biofilm for the removal of cobalt, strontium and caesium from aqueous solutions. Hydrometallurgy, vol. 83, Issues 1–4, 2006, p. 141–145.

Speaker: Prof. Liuba Coretchi (National Centre of Public Health, Academy of Sciensis of Moldova)

09:00 PROBLEMS OF REHABILITATION OF URANIUM WASTE IN MOUNTAIN AREARS OF KYRGYZSTAN

Abstract: Mountain - is the last major regions where is the last natural landscapes and there are used in traditional mode (nomadic, semi-nomadic herding, etc.). The extraction and transportation of minerals are one of the major threats to biodiversity of mountain ecosystems in the country. After (1991) independence in the country there is begun an intensive exploration of mineral resources and environmental issues and the environment remain in the background. If the ecosystem of the mountain will be destroyed, then disappear related components of biodiversity (the richness of the country, including endemic and others), the geo-morphological processes will active in the mountains (landslides, avalanches, erosion of soil and rock, etc.). In some areas of Kyrgyzstan over the past 80 years of mining has accumulated vast amounts of structures in the form of waste dumps, tailings ponds and dross ponds. On the territory of Kyrgyzstan was formed 55 tailings, the total area of 770 hectares, of which more than 132 made about million m3 of tailings and 85 mining dumps are accumulated 700 m3 waste, cover an area of over 1,500 hectares, of which 31 and 25 tailing dumps - uranium waste, the volume of 51.83 million m3. As period of 2012, their total radioactivity was over than 90 thousand curies [1, 2, 4]. It is known that almost any interior design involves direct destruction of natural ecosystems. Mines, quarries, dumps, and the crowded river channels allocated to significantly alter the landscape, causing ecosystem changes in the scale of entire watersheds and river basins. The network infrastructure of the mining industry (roads, railways, gas and oil pipelines, power lines and other communications, shift camps, etc.) entangles more extensive areas, mountains and fragmenting populations of vulnerable species, becoming a constant threat to wild animals and plants. The extraction and primary refinement of fossil raw materials are often a powerful source of air pollution and natural waters, the bane of all sorts (including the highly toxic heavy metals, radionuclides, dioxins, cyanides, etc.), dust (also often poisoned), as well as noise and thermal pollution, even in normal mode. But in this production are frequent accidents and disasters. Many of the tailings in the country are located in populated areas, seismic and landslide prone mountain areas. Not yet fully known radiological impact on the environment and the population - low standard of living, social and migration issues, etc. contribute to the overall poor socio-psychological situation in these areas, including threats and risks from radiation and other potential is physical risks. For example, the uranium tailings and waste damp of Biosphere territories of Issyk-Kul is located 2.5 km east of the residential village, but due to natural factors (rainfall, groundwater, landslides and mudflows) an environmental threat to the Issyk-Kul Lake (1.5 km from the lake) and the nearest villages, located on the slopes with a slope between the mountains to 30-45 ° [1, 2]. Present time the protective structures and some sections of the surface of uranium tailings and waste dumps are destroyed without untimely repairs and services in the country. It was established that in mountain arears of the uranium tailings as "Tuyuk Suu", "Taldy-Bulak" and some areas of uranium technogenic province “Mailu-Suukoy” are potentially dangerous, which are included in the register of the Ministry of Emergency Situations of the Kyrgyz Republic. It should be emphasized the fragility of the soil in the mountainous areas of biosphere in Kyrgyzstan. It has a low power (20 to 50 cm), and its preservation depends on the ecological balance in the mountains (especially mining development, etc.) and the unpredictable consequences of natural and manmade disasters in the surrounding area and the region. It should be noted that the radiological analysis of the tail materials and assessment of the state of the environmental contamination with radionuclides are not done after the close of many uranium tailings and waste dumps in the Republic, we should be continue the work in this direction of research and monitoring [3, 5]. REFERENCES [1] Djenbaev B. M., Mursaliev of A.M. Biogeochemistry of natural and technogenic ecosystems of Kyrgyzstan. – B.:Science, 2012, 404 p. [2] Djenbaev B.M., Kaldybaev B.K., Zholboldiev B.T. - Book title: Radioactive Waste (ISBN 978-953-51-0551-0). 502 pages, April, 2012. Book edited by: Dr. R. O. Abdel Rahman (http://www.intechopen.com/articles/show/title/) [3] Standards of radiation safety (NRB-99). 2.6.1.758-99. [4] Torgoyev I.A. Geoecology and waste of a mining complex of Kyrgyzstan. Bishkek:Science. 2009. 240 p. [5] International atomic energy agency, Quantification of radionuclide transfers in terrestrial and freshwater environments for radiological assessments, IAEA-TECDOC-1616, Vienna: IAEA, 2009. – 51 p.

Speaker: Prof. Bekmamat Djenbaev (Kaldibaev B.K)

09:00 SOME METALS CONCENTRATION, INCLUDING RADIOACTIVITY, IN WATER AND SOIL SAMPLES AT SOME LOCATIONS NEAR THE HOTMUD FLOW AT PORONG DISASTER AREA, SIDOARJO, EAST JAVA, INDONESIA.

More than hundred thousands people in the Porong subdistrict have been displaced by the hot mud flowing from a natural gas well being drilled by Lapindo Brantas, an oil well company since late May 2006. The mud was estimated to be flowing at a rate of 125,000 m3 per day and rapidly flooded surrounding areas displacing hundred thousands of people. For about thousand people had to seek medical treatment after exposure to and inhalation of a poisonous gas. This disaster is happen continue until now. The monitoring of mud flow and water quality are performed by taking separate samples in the field and transporting the samples to the laboratory for analysis. The sediment and water samples of hot mud flow at Porong Area, East Java, Indonesia, were taken from several points and analyzed are as follow ; concentration of Cr, Cd, Zn, Cu, Pb, Fe in sediment samples in the range of : 0.16 – 0.20; 0.16 – 0.34; 54.80 – 61.10; 21.22 – 24.16; 3.45 – 9.83; 1314 –1526 ppm respectively. And concentration of Cr, Cd, Zn, Cu, Pb, Fe in water sampels in the range of : 0.044 – 0.072; 0.001 – 0.002; 0.19 – 0.42; 0.012 – 0.036; 0.26 – 0.36; 0.20 – 0.58 ppm respectively. The soil and water samples were taken from several locations near the hot mud flow at Porong Area, East Java, Indonesia and the concentration of Cr, Cd, Zn, Cu, Pb, Fe in soil and water samples are not detected. Keywords : monitoring, metals, sediment, water, hot mud flow, Porong Area Indonesia

Speaker: Dr adam wiryawan (department of chemistry, brawijaya university)

09:00 STUDY ON ENVIRONMENTAL REMEDIATION AND RADIOACTIVE WASTE (NORM) MANAGEMENT FROM THE RARE-EARTH RESEARCH & DEVELOPMENT CENTER IN THAILAND

STUDY ON ENVIRONMENTAL REMEDIATION AND RADIOACTIVE WASTE (NORM) MANAGEMENT FROM THE RARE-EARTH RESEARCH & DEVELOPMENT CENTER IN THAILAND YA-ANANT, Nanthavan, KASEMTANASAK, Virut, NUANJAN, Panya, PATTANASUB, Archara, AKHARAWUTCHCHAYANON, Thunyaras, SRIMORK, Panuwat, PRASERTCHIEWCHAN, Nikom and PRUANTONSAI, Paphot Radioactive Waste Management Center, Thailand Institute of Nuclear Technology (TINT), 16 Vibhavadi-Rangsit Road, Bangkok, Thailand. E-mail address: nanthavan@tint.or.th Abstract: The rare-earth extraction and the U-Th extraction pilot plans from the Rare-earth Research and Development Center at Pathumthani province, Thailand was operated during 1995-2005 after that the activities were no longer in operation. After the big flood occurred in 2011, the pilot plants and areas around were unfortunately contaminated with radioactive materials (NORM). This paper introduces the survey procedures for checking contamination of the rare-earth extraction and the U-Th extraction pilot plans for the further decontamination, and radioactive waste management.The environmental samples, such as sludge, waste water, aquatic plants, and fishes around the site are investigated. The radiological parameters under investigation are gross-alpha, gross-beta, and radionuclides from soil, water, aquatic samples on-site the plants. Radioactivity level compared with the base-line radioactivity level before the operation of the plants (1988-1985) were found relatively high. The results confirmed that the site was contaminated by residual radioactive material. The environmental remediation is need for the safety aspects.

Speaker: Mrs NANTHAVAN YA-ANANT (Thailand Institute of Nuclear Technology)

09:00 ОБЗОР РАДИАЦИОННЫХ И ЭКОЛОГИЧЕСКИХ ПРОБЛЕМ В РАЙОНЕ РАЗМЕЩЕНИЯ ХВОСТОХРАНИЛИЩ МАЙЛУУ-СУУ, МИНКУШ И КАДЖИ-САЙ, КЫРГЫЗСТАН

Кыргызская Республика находится во юго-востоке Центральной Азии, граничит с Казахстаном, Таджикистаном, Узбекистаном и Китаем. На территории Кыргызской Республики размещено 35 хвостохранилищ и 37 отвалов горных пород, содержащих радиоактивные отходы. В последние годы в Кыргызской Республике активно осуществляются работы по ремедиации уранового наследия, заключающиеся в инженерных исследованиях, характеризации отходов, разработке проектов и экологических оценок. С 2005 г по 2013 г на участке осуществлялись ремедиационные действия на хвостохранилищах и отвалах рядом с г. Майлуу-Суу. Главным обоснованием для начала восстановительных мероприятий послужили геомеханические риски разрушения хранилищ с отходами, расположенными рядом с рекой Майлуу-Суу, и риски трансграничного загрязнения окружающей среды. Проектирование и финансирование ремедиационных действий для объектов уранового наследия в г. Майлуу-Суу, пгт. Минкуш и пгт. Каджи-Сай в Кыргызской Республике осуществляется с помощью Всемироного Банка, Еврокомиссии и Правительства Российской Федерации. Задачей для регулирующих органов и операторов, задействованных в ремедиационных действиях на объектах уранового наследия, оценить эффективность уже предпринятых действий, оценить уровень угроз, исходящих от радиоактивных отходов, и начать разработку стратегии по внедрению институционального контроля за объектами. Институциональный контроль является новым термином для молодой республики, чьи усилия существенно ограничены экономическим кризисом. Большинство законодательных актов Кыргызской Республики носят рамочный характер, излишне абстрактны и декларативны, отсутствуют нормы прямого действия, что делает необходимым конкретизацию процедур и механизмов их исполнения в подзаконных актах. В настоящее время в Кыргызской Республике осуществляются различные программы по развитию подзаконных актов, направленных на обеспечение экологической и радиационной безопасности при обращении с радиоактивными отходами и при проведении ремедиационных действий. На хранилищах участков в Майлуу-Суу, Минкуш и Каджи-Сай в Кыргызской Республике размещено более 5,2 млн.м3 хвостовых и рудных отходов бывшего уранового производства. Все объекты размещены в горной местности, на берегу водных объектов и рядом с населенными пунктами. Основной вклад в загрязнение окружающей среды в районе размещения объектов уранового наследия дают инфильтрация сульфатов, радионуклидов и тяжелых металлов в подземные воды, разрушение боковых склонов хранилищ и поверхности хвостохранилищ. На состояние всех этих объектов отрицательно повлияли несколько факторов, основные из них: политический (отсутствие стратегии, неэффективные регулирование и несовершенная законодательно-нормативная база), временной (жизненный цикл для большинства из искусственных объектов составляет более полувека) и человеческий (недостаток профессионально обученных специалистов). В населенных пунктах рядом с объектами уранового наследия отмечаются случаи повышенного содержания радона-222 в воздухе и повышенной мощности эффективной дозы гамма-излучения в жилых домах. Загрязнение подземных и поверхностных вод является нежелательным явлением, не только для Кыргызской Республики, но и в трансграничном аспекте, так как горные реки Кыргызской Республики имеют свое продолжение в соседних странах. Изучение исторической информации о хвостохранилищах уранового наследия, современная характеризация объектов уранового наследия, планы ремедиации и уже выполненные восстановительные мероприятия позволяют оценить эффективность ремедиационных действий, найти интституциональные и регулирующие пробелы в инфраструктуре Кыргызской Республики и разработать стратегию по развитию безопасного обращения с хвостохранилищами и отвалами уранового наследия. REFERENCES [1] Final Report (NATO – 2009-2010) NATO SfP Project 981742 (RESCA) «Uranium extraction and environmental security in Central Asia» [2] End Of Mission Reports (IAEA/NATO-2008) KIG 9004 01: IAEA Technical Co-operation Expert Mission «Support for radiological field studies at Minkush» and NATO SfP Project 981742 (RESCA) [3] Карпачев Б., Менг С. «Радиационно-экологические исследования в Кыргызстане», 2000 г.

Speaker: Mr Muratbek Kalykov (author)

11:00 → 11:30 Coffee break

11:30 → 13:30 Session 4A - 2: Technical and Technological Aspects of Implementing Decommissioning Porgrammes - Parallel Session

The purpose of this session is:
• To review progress in decommissioning technologies and cost estimation over the past decade and to identify challenges and needs in the future
• To review current developments, including from damaged facilities, with a view to identify existing gaps in knowledge and needed improvements

11:30 Selection of strategy and technology for segmentation of pressure vessel and internals at Zorita NPP

In 1964, construction started on the José Cabrera Nuclear Power Plant, located in Almonacid de Zorita, 43 miles east of Madrid, Spain. The José Cabrera Nuclear Power Plant is a single-loop pressurized water reactor with 160 MWe of power installed. The main components, particularly the vessel and vessel internals, were manufactured and supplied by Westinghouse. The plant operated between 1968 and 2006. It is the first light water reactor being dismantled in Spain. In mid-2010, the Empresa Nacional de Residuos Radiactivos (ENRESA), the Spanish Radioactive Waste Management Agency, within the works of the Plan de Desmantelamiento y Clausura (Dismantling and Closure Plan) for the José Cabrera Nuclear Power Plant, awarded the segmentation of the vessel internals at Zorita, a project that was finalized in fall of 2013. ENRESA awarded the segmentation work for the reactor vessel in September 2013, and the work was completed in the beginning of April 2015. The underwater mechanical cutting technology used in the segmentation of the reactor internals demonstrated safety and efficacy in cutting materials that are stronger and harder than those of the vessel, and with more complex geometries. Same technology was used for cutting the vessel internals as well. The main conclusions and lessons learned that ENRESA has derived from these projects are: • Mechanical cutting has proven effective for cutting complex geometry and radiation induced hardness components. • Although slightly slower than other alternatives, this technology has shown many advantages, such as: easy implementation, very low radiological burden and much reduced generation of secondary wastes. • The decision to use fuel-type containers, with modified canisters, for the interim storage of the non-LILW primary and secondary wastes, has been proven as a straightforward solution for their management. • The reuse of some of the equipment utilized previously for the segmentation of the reactor internals helped the vessel segmentation project to be also cost effective and time saving. • The use of CE-2B baskets for interim storage in the Reactor Cavity greatly facilitated the waste packaging operations, making them independent of the cutting sequence. However, this required an extensive effort of identification and tracking of individual pieces. • The use of “Inserts” greatly facilitated the final non-LILW Waste Container (GWC) loading, also allowing the optimization of weight and activity distribution, at GWC level and between them. • The use of the “Chip Boxes” for secondary waste was shown to be an effective way to contain and dry them, in the case of those introduced in the GWC. • Special attention has been drawn at minimization of the collective dose and reduction of the waste volume via an optimum pattern cutting • Visibility problems were solved with the installation of an additional high flow water filtration system.

Speaker: Mr Manuel Manuel Ondaro Pino (ENRESA)

11:30 → 13:30 Session 4B - 2: Technical and Technological Aspects of Implemeting Envrionmental Remediation Programmes - Parallel Session

The purpose of this session is:
• To review the available technologies for environmental remediation
• To define the existing gaps in knowledge and needed improvements with the aim of facilitating the implementation these activities.

11:30 Long-term water treatment at uranium mining sites

Water treatment at active or closed uranium mining sites is required whenever the effluent from a mine, a waste depository or any other entity reveals higher concentrations than permitted under the respective Water Act or with respect to dose calculations according to the radiation protection regulations. Decision criteria for the selection of the optimal treatment approach are site-specific and include (a) type and concentrations of the contaminants to be removed, (b) possibility or requirement to recover certain components, (c) flow rate and contaminant loads to be removed, (d) time frame at which water treatment is needed, (e) fluctuations of water composition, contaminant load and/or flow rate, (f) type and amount of residues produced. The list of ‘best available technologies economically achievable’ (BATEA) has been collected most recently for the Canadian mining industry (Hatch 2014). The most commonly used active treatment process worldwide is lime treatment. This technology produces contaminant-laden iron and/or gypsum-rich sludge which has to be properly managed and disposed. Most lime operations internationally are HDS (high density sludge) plants. Yet there are several locations with other treatment technologies, such as ion exchange or reverse osmosis, but their application is restricted to localized cases. Besides the minimization of emissions, preferred mine water treatment concepts are intended to recover the mineralization loads not as waste but rather as valuables, such as uranium, metals, fertilizers or sulphur. Passive treatment is often the choice for low flow or low pollution waters and for remote abandoned mine sites, however, those options are not applicable for all types of mine water, and treatment efficiency can be limited due to seasonal changes in the systems parameters. Water treatment is a crucial activity in the context of the remediation at uranium legacy sites. One of the most prominent projects of that kind is the Wismut Environmental Remediation Program in Germany. This program is focused on the remediation of mining legacies at several former production complexes which were operational between 1946 and 1990. While all mines and mills have been decommissioned and physical remedial work is to a large extent complete, long-term water management will continue to require the provisions of considerable funds. In order to ensure compliance with protection goals for receiving streams and aquifers Wismut GmbH currently operates six water treatment plants (WTP) and the associated systems for water catchment, delivery and discharge as well as for the conditioning and disposal of treatment residues. Total annual water treatment throughput during the period 2010 - 2014 amounted to approximately 20 million m³ of contaminated waters. At all WTPs currently in operation, the separation of key contaminants such as uranium, arsenic, radium-226 as well as of iron, manganese and other heavy metals involves precipitation technologies in various modifications. Only at the former Königstein underground ISR operation site uranium is still separately recovered. Management of mine waters requiring treatment is the number one core task of the long-term water management activities. The ongoing retrofitting of WISMUT’s 1st generation treatment plants relies on one or several of the following factors: (a) changes in quantity and/or quality of waters to be treated, (b) cost optimization, (c) tightening of treatment standards, (d) advances on state of the art technology. Smaller satellite WTP’s are now being operated by remote control and remote surveillance. Projected technological adaptions in existing facilities are aimed first and foremost at improving uranium removal and optimizing residue disposal. Process combinations of ion exchange and fixed bed adsorption are under consideration as alternative technology. Technologies for sulphate removal are primarily triggered by the need to reducing neutral salt loads. Newest research focuses on membrane-based techniques in combination with precipitation and/or evaporation of retentates. Other key activities comprise the process optimization regarding energy consumption. In addition, the utilization of renewable energies is also under review, such as the use of the geothermal potential of mine waters for heat supply. One of the sustainability goals of the Wismut remediation program stipulates that remedial operations shall lead to a self-sustaining system status which will dispend with active measures in the long term and carry low residual risks. To achieve this, extensive research is undertaken to enhance natural attenuation processes in mining influenced water bodies. For instance, a technology to inject alkaline and reducing solutions was developed for the remediation of the Königstein mine and successfully tested in a field test. Another potential technological approach is based on stimulating a microbially catalyzed autotrophic reduction-oxidation process chain within a reactive zone. Prospects to initiate and sustain a microbially catalyzed reductive uranium immobilization are indeed realistic as is impressively shown at the Pöhla mine: Spontaneous reductive uranium precipitation produces uranium concentration levels of < 50 µg/l in emerging flood water for almost two decades.

Speaker: Dr Michael Paul (Wismut GmbH)

12:30 Review conditions of the environment in the placement of tailings in Kara-Balta town, Kyrgyzstan. and methods of development strategy of the remediation plan for this tailings pounding

IAEA INTERNATIONAL CONFERENCE ON ADVANCING THE GLOBAL IMPLEMENTATION OF DECOMMISSIONING AND ENVIRONMENTAL REMEDIATION PROGRAMMES (MAY 2016) ОБЗОР СОСТОЯНИЯ ОКРУЖАЮЩЕЙ СРЕДЫ В РАЙОНЕ РАЗМЕЩЕНИЯ ХВОСТОХРАНИЛИЩА В Г. КАРА-БАЛТА, КЫРГЫЗСТАН. И СПОСОБЫ РАЗВИТИЯ СТРАТЕГИИ ПЛАНА РЕМЕДИАЦИИ ДЛЯ ЭТОГО ХВОСТОХРАНИЛИЩА Solomatina, Anna, Decommissioning and Remediation Unit, Waste and Environmental Safety Section, Address and Street, Kara-Balta town, Kyrgyzstan E-mail address: Solomatina_76@mail.ru Abstract: Территория Кыргызской Республики в прошлом веке была одним из основных источников природного урана в СССР. Кыргызская Республика находится во юго-востоке Центральной Азии, граничит с Казахстаном, Таджикистаном, Узбекистаном и Китаем. На ее территории размещено 35 хвостохранилищ и 37 отвалов горных пород, содержащих радиоактивные отходы. После развала СССР Кыргызская Республика, находясь в условиях экономического кризиса, встала перед необходимостью поддерживать безопасность окружающей среды вокруг хвостохранилищ. В последние годы в Кыргызской Республике осуществляются работы по улучшению системы мониторинга, политики и стратегии по управлению радиоактивными отходами, активно осуществляются работы по ремедиации уранового наследия. Для еще действующих хранилищ радиоактивных отходов урановой и цветной промышленности Кыргызской Республики необходимо разработать стратегию по эффективной защите окружающей среды и населения, а также план ремедиации. В настоящее время ни одно из действующих предприятий Кыргызской Республики, где образуются радиоактивные отходы, не имеет собственного плана ремедиации и денежного фонда для развития восстановительных мероприятий. Для внедрения стратегии защиты окружающей среды и развития планов для ремедиации в Кыргызской Республике в настоящее время, кроме инженерных и экологических изысканий на хвостохранилищах, укрепления институциональной базы и развития законодательной базы, осуществляется изучение опыта прошлых лет, имеющегося в Кыргызской Республике на примере хвостохранилища в г.Кара-Балта. Ситуация вокруг хвостохранилища в г. Кара-Балта с точки зрения экологической и радиационной безопасности является интересной задачей для регулирующих органов и операторов, так как в жестких экономических условиях требуется найти консенсус для сохранения оптимальных условий для площадки 1. INTRODUCTION Хранилище хвостовых отходов уранового производства г. Кара-Балта является крупнейшим хранилищем низко радиоактивных отходов в Кыргызской Республике, на площади в 256 га размещено 76,8 % от общего объема радиоактивных хвостовых отходов, имеющихся в Кыргызской Республике. Радиоактивные отходы, образуются на гидрометаллургическом производстве, принадлежащем Кара-Балтинскому горно-рудному комбинату, основанному в 1955 г. В настоящее время предприятие ориентировано на выпуск только закиси-окиси урана. На хвостохранилище хранятся остатки переработки урановой руды и отходы, образующиеся в результате экстракции химических урановых концентратов, общим объемом 37,1 млн. м3. 2. METHDOS AND RESULTS Для данной площадки имеется большое количество данных, характеризующих хранящихся там радиоактивных и не радиоактивных отходов, состояние окружающей среды и радиационного фона. Используя имеющуюся базу, необходимо развить стратегию плана ремедиации для хвостохранилища. Для усиления регулирующих механизмов по ремедиации и привлечения внимания оператора к данному процессу предложить стратегию развития плана ремедиации в качестве пилотного проекта в республике. Состояние окружающей среды вокруг хвостохранилища неоднородно по всей территории, где размещены площадка с отходами, санитарная зона и населенный пункт [1], [2], [3]. Основным реципиентом загрязнения в районе хвостохранилища являются подземные воды, ореол загрязнения подземных вод в настоящее время составляет 4 км2. Основными загрязняющими веществами в водах являются сульфаты, аммоний, нитраты, железо и некоторые тяжелые металлы, такие как марганец и молибден. В истории хвостохранилища содержатся разные факты: аварийный перелив хвостовых отходов в сторону жилой застройки, реконструкция дна секций хвостохранилища, перепрофилирование предприятия с переработки рудного сырья на переработку химических концентратов с неизбежным снижением объемов, эксперимент по очистке подземных вод с помощью сульфатредуцирующих бактерий [5], комплексные изыскания [1] ], [2], [3]. 3. CONCLUSIONS Изучение исторической информации о хвостохранилище в г. Кара-Балта уже сейчас позволяет разработать стратегию по развитию плана ремедиации этого хвостохранилища. REFERENCES [1] Радиоэкологические и смежные проблемы уранового производства, часть 2, Национальная академия наук Кыргызской Республики, Бишкек (2004). [2] Отчет "Экологический мониторинг и развитие потенциала управления» ТА № 2934 КGZ", Бишкек (2004). [3] Комплексный отчет по изучению состояния окружающей среды в зоне влияния Кара-Балтинсского горно-рудного комбината, Кара-Балта, (2005 г) [4] Отчет "Проведение исследований по очистке загрязненных подземных вод на хвостохранилище Кара-Балта с применением способа микробиологической очистки, ВНИПИПРОМТехнологии, Москва (1997) [5] Отчет по оценке угроз, Collaborative Project of the Norwegian Radiation Protection Authority and SAEP (Kyrgyz Republic «Поддержка в развитии регулирующего органа по радиационной и ядерной безопасности в Кыргызской республике», Бишкек (2012).

Speaker: Ms Anna Solomatina (Chui ecological laboratory)

13:00 The NEA Thermochemical Database Project: 30 Years of Accomplishments

The NEA Thermochemical Database (TDB) Project (www.oecd-nea.org/dbtdb/) is a joint project of the Radioactive Waste Management Committee and the Databank that was launched in 1984 [1]. It consists of a database of chemical thermodynamic values treating the most significant elements related to nuclear waste management, that fills significant gaps in radionuclide chemistry and supports the modeling requirements for performance assessments of radioactive waste disposal systems. The work carried out since the initiation of the project has resulted in the publication of thirteen major reviews and a large set of selected values that have become an international reference in the field.

Speaker: Ms MARIA ELENI RAGOUSSI (OECD NEA)

13:30 → 15:00 Lunch break

15:00 → 16:30 Session 4A - 3: Technical and Technological Aspects of Implementing Decommissioning Programmes - Parallel Session

15:00 DECOMMISSIONING OF KOZLODUY NPP UNITS 1 TO 4 - PROGRESS AND CHALLENGES

KAZAKOV, Momchil, Specialised Division Decommissioning Unit 1-4, SE RAW Address: Bulgaria, Kozloduy, Kozloduy NPP Site Email-address: momchil.kazakov@ie.dprao.bg 1. Introduction: State Enterprise Radioactive Waste – National Operator for the Activities for RAW Management and Decommissioning of Nuclear Facilities –background, management, structure. Kozloduy NPP Units 1-4, history of the units operation and shutdown. Normative framework. Funding. 2. The presentation: • Strategy for Decommissioning of Units 1-4 of Kozloduy NPP – Decommissioning Plan for KNPP Units 1-4, terms, key phases, final objective – brown field. • Dismantling activities – Dismantling of equipment in Turbine Hall of Power Units 1-4 – amounts planned, cutting techniques, results, pictures. Activities in the Controlled Area – radiological survey, decontamination, preparation for dismantling activities. Figure 1.Dismantling of TG–3 Rotor of Unit 2 • Free release of dismantled materials – Normative framework, Criteria, Free release procedures for different equipment type and various materials. • Projects implementation – Projects for RAW treatment – solidified phase from evaporator concentrate tanks (ECTs), spent ion-exchange resins, sludges and solid RAW; Supply of different types of containers for transport and storage of materials; Sites for materials management; Supply of lifting and transportation equipment; Design and construction of Size Reduction and Decontamination Workshop; Free release measurement facility. Figure.2 Facility for Retrieval and Stabilization of Spent Ion Exchange Resins - principle scheme. • Important steps until 2015 – Receipt of decommissioning licence; Programmes; Procedures; Demolition of civil structures in Turbine Hall of Power Units 1 and 2; Dismantling of equipment inTurbine Hall of Power Units 1-4; Free released materials; Activities in the Controlled area – relocation of the boron absorber rods and dummy assemblies from the four reactors to the Spent Fuel Pool (SFP-3) with the purpose to reduce the dose budget of the personnel; Schedules, performance, results. Figure.3 Preparation for removal of dummy assemblies, Unit 2, February 2015 • Activities planned – Commissioning of RAW treatment facilities; Commissioning of Size Reduction and Decontamination Workshop; Decontamination of the main equipment and facilities for personal dose reduction; Dismantling of Controlled Area equipment; Management of Radioactive materials and RAW. 3. Conclusions: Performance Analysis, expected difficulties, suggested solutions.

Speaker: Mr Momchil Kazakov (State Enterprise Radioactive waste, Bulgaria)

15:05 ISDC Approach to Probabilistic Cost Risk Assessment

ISDC Approach to Probabilistic Cost Risk Assessment Vladimir Daniska(1), Thomas LaGuardia(2), Helena Daniskova(3), (1)DECOM, a.s., Trnava, Slovakia; (2)Thomas LaGuardia & Associates, USA, (3)Comenius University, Bratislava, Slovakia; Abstract The existing ISDC cost format [1] was developed based on results of deterministic costing methods; the resulting estimate represents the best judgement by the estimator of the total expected cost, based on the boundary conditions established in the ‘basis of estimate’ document. On this basis, the project budget established in accordance with the ISDC is fully expected to be spent during project implementation. The increasing use of probabilistic methods to consider the cost impact of risks, i.e. scenarios and discrete events not necessarily included with the basis of estimate will require that the ISDC cost presentation formats will need to be extended. In line with this approach, the project owner may establish a project budget that goes beyond the best judgement of the total expected cost, allowing a margin to address costs associated with risks events beyond the defined project scope. The degree of margin provided by the budget is a reflection of the ‘risk appetite’ of the project owner or responsible funding organisation. This poster presents options for extension of the ISDC cost presentation format to take account of additional costs determined by probabilistic approaches. The proposed format includes results of sensitivity calculations, undertaken to determine the main cost drivers, and to determine which activities should be the subject of Monte Carlo analysis. It also includes an approach to presentation of the results that allows the funder to decide what level of risk to accept. [1] OECD NUCLEAR ENERGY AGENCY, INTERNATIONAL ATOMIC ENERGY AGENCY, EUROPEAN COMMISSION, ‘International Structure for Decommissioning Costing (ISDC) of Nuclear Installations’ OECD/NEA No. 7088, (2012)ISDC.

Speaker: Dr Vladimir Daniska (DECOM, a.s.)

15:10 INTEGRATED 3D SUPPORT SYSTEM FOR IMPROVING SAFETY AND COST-EFFICIENCY OF NUCLEAR DECOMMISSIONING PROJECTS

Abstract: A significant number of nuclear power plants will have to be decommissioned over the next few years as a result of earlier and planned shut downs initiated by plant aging, political decisions and unfortunate events. The decommissioning process is challenging for all stakeholders due to uncertainties and risk associated with decisions on applied technologies, organisational changes, and management of human factors. In this study we investigated how concepts, enabled by advanced information technologies, can be applied for providing continuity between different phases of the decommissioning work process and life cycle of the installation, as well as stakeholders involved in on-going and future decommissioning projects. 1. INTRODUCTION With the emergence and exploitation of alternative energy sources, and unfortunate events impacting public confidence in safety, justifying the use of nuclear energy for electricity production is increasingly important. Due to the efforts involved to ensure that the work is done safely and the possibility of long term waste storage or site control commitments, decommissioning entails high costs influencing investment per unit energy required from the energy producers. For nuclear energy production to remain sustainable, technologies enabling more efficient methods for developing optimal work strategies and prepare personnel and the organisation for efficient implementation of the decommissioning plans are essential. 2. METHDOS The scope of this activity is to investigate how new methods enabled by emerging information technologies like 3D real-time radiological and work simulation (Figure 1), advanced user interfaces, and mobile computing, can be utilised for optimising safety and costs in the decommissioning of nuclear installations. 3. RESULTS Research results and lessons learned from industrial applications shows that new concepts enabled by emerging 3D computing technologies have great potential for improving nuclear decommissioning strategies. Such techniques offer very effective new opportunities for improving early characterisation and strategical decision making, knowledge management, on-site management of radiological waste, briefing and training of field workers, and regulatory compliance. Figure 1: Figure 1. User interface of the VRdose system, an interactive work planning tool with real-time 3D radiological modelling and visualisation capabilities. 4. CONCLUSIONS 3D modelling has become an essential tool in the design and licencing of new nuclear installations. In addition, our earlier research results demonstrate that 3D simulation can be efficiently utilised for improving safety and efficiency of maintenance jobs. An increasing number of organisations are adopting 3D simulation technology to support the decommissioning of aged nuclear installations. However full scale international investment and application of new methods enabled by such technology requires research results describing good practices, as well as proving safety and cost benefits. REFERENCES [1] SZŐKE, I., et al., “New software tools for dynamic radiological characterisation and monitoring in nuclear sites” in: Proceedings of Workshop on Radiological Characterisation for Decommissioning, Nyköping – Sweden, April 17-19 (2012) http://www.oecd-nea.org/rwm/wpdd/rcd-workshop/ [2] SZŐKE I., JOHNSEN T., “Human-centred radiological software techniques supporting improved nuclear safety” Nuclear Safety and Simulation, 4(2013) 219-25. http://www.ijnsweb.com/?type=subscriber&action=articleinfo&id=176 [3] SZŐKE I., et al., “Real-time 3D radiation risk assessment supporting simulation of work in nuclear environments” Journal of Radiological Protection; 34(2014) 389–416. http://iopscience.iop.org/0952-4746/34/2/389/ [4] SZŐKE I., et al. “Comprehensive support for nuclear decommissioning based on 3D simulation and advanced user interface technologies” Journal of Nuclear Science and Technology, 52(3), 371-82, 2014. http://www.tandfonline.com/doi/full/10.1080/00223131.2014.951704 [5] CHIZHOV K., SNEVE M., SZŐKE I., et al. “3D simulation as a tool for improving safety culture during the remediation work in the Andreeva Bay” Journal of Radiological Protection.; 34(2014) 755-73,. http://iopscience.iop.org/0952-4746/34/4/755/ (Bernard Wheatley Award for 2014)

Speaker: Dr István Szőke (Institute for Energy Technology)

15:15 R&D Outline for Decommissioning of the Fukushima Daiichi Nuclear Power Station

Abstract International Research Institute for Nuclear Decommissioning ( IRID) was established in August 2013 in Japan as an organization to develop nuclear power plant decommissioning technologies efficiently. The main purposes are at first technology development for nuclear decommissioning, secondly, promoting cooperation with international and domestic organizations on nuclear decommissioning, and thirdly developing human resources for research and development. Since then, IRID has been playing a proactive role in the R&D required for the decommissioning of Fukushima Daiichi Nuclear Power Station of TEPCO in Japan as an urgent issue. In this poster session, I will introduce some examples of IRID’s R&D activities for preparation of fuel debris retrieval that is a core operation of decommissioning. Various kinds of remote controlled equipment and robots have been developed so far for decontamination and investigation inside the reactor building. In 2012 and 2013, we investigated the dose rate and contamination distribution at each floor of the Units 1-3. Therefore, the conditions inside the reactor buildings are still very severe. We have developed three types of remote decontamination equipment: suction/blast type, high pressure water jet type and dry ice blast type. As the most recent example, IRID developed a shape changing robot that can go through a penetration to investigate outside the pedestal at the lower part of the RPV. IRID also developed a technology for detection of fuel debris in the reactor. Remote sensing technology utilizing cosmic ray muon is one of the methods to identify location of fuel debris. In the development of technologies for fuel debris retrieval, in addition to the method in which PCV is submerged, we are evaluating retrieval in the air, partial or full in air, as applicable method. Because the status differs from unit to unit, we should consider the applicability of each method. IRID is also studying the technologies to establish concept for treatment and disposal of accident generated waste. Radionuclide analysis of rubbles, fallen trees and contaminated water, etc. sampled at Fukushima Daiichi NPS and inventory evaluation of waste materials based on the analysis of these results are now being performed. In order to have a clear prospect of safe treatment and disposal of solid waste, IRID will continue to conduct R&D of technologies required for storage management, waste characterization, waste encapsulation, and waste disposal. As the result of our R&D activities, IRID has acquired some useful outcome, but at the same time, technical challenges toward decommissioning have also becoming clearer. Based on these achievements and challenges, IRID will keep working on technology development necessary to decide the method for fuel debris retrieval in 2018, and contributing to completion of decommissioning at the earliest time.

Speaker: Mr Hideaki Ohhashi (IRID)

15:20 RECENT IMPROVEMENT AND LESSONS LEARNED USING IMAGING TOOLS FOR RADIOLOGICAL CHARACTERIZATION

During decommissioning and dismantling operations of Nuclear Plants, imaging devices allow a fast and accurate identification of contaminated areas. Since more than twenty years, CEA is developing such cameras to localize hot spots in high and low level dose rate environment. One of the most famous tools is the “gamma camera” dedicated to the localization of gamma sources from 50 keV to 1500 keV. They have been extensively qualified in various plants such as experimental reactors or reprocessing plants, they also have found applications for waste management and hot cells measurements. More recently, the CEA has developed a complementary imaging device dedicated to the alpha contamination detection. This imaging measurement is based on the detection of UV radiation emitted by nitrogen subjected to alpha particles irradiation. Regular alpha contaminated area measurements have to be carried out at contact of the surface due to the low path of alpha particles in air (~ 4cm for  from 241Am source). This new technique allows building images of the sources at distance and through translucent materials. This alpha camera has been tested on site and is currently under industrialization process. The very last studies concern the development of a dual alpha/gamma compact camera based on the feedback of both Aladin gamma camera and the alpha camera. This prototype has been patented in 2015. Finally, The CEA is also qualifying and developing new field of portable gamma camera with low weight. 1. INTRODUCTION The knowledge of the radiological state of a process or a facility is of prime importance not only during the initial stages of a dismantling project’s initial inventory, but also during the follow-up phases of clean-up and final checks. During a facility’s operation, a clear view of the process radiological level is equally necessary, in order to plan maintenance scheduling and optimize interventions by personnel. Radiation protection teams generally ensure worker radiological safety, and also supply dose mappings for each of the areas associated with typical spectra. In most cases, this mapping does not give enough information to be used as input data when preparing maintenance or clean-up work. With advances in activity modeling calculation codes (using 3D), the way measurements are made to recover the best input radiological data on site is a key factor. 2. LESSONS LEARNED FROM USING IMAGING TOOLS The CEA has developed many compact characterization tools to follow sensitive operations in a nuclear environment. Usually, these devices are made to carry out radiological inventories, to prepare nuclear interventions or to supervise some special operations. These in situ measurement techniques mainly take place at different stages of clean-up operations and decommissioning projects, but they are also in use to supervise sensitive operations when the nuclear plant is still operating. In addition to this, such tools are often associated with robots to access very highly radioactive areas, and thus can be used in accident situations. The CEA has also carried out more than several hundred radiological investigation using imaging devices like gamma cameras and alpha cameras. 3. RECENT ADVANCES OF IMAGING TOOLS The CEA is currently studying the possibility to use only one compact device to measure both alpha contamination and gamma irradiation. As a matter of fact, some nuclear facilities under decommissioning operations have to face these two issues at the same time. A first prototype has been developed and tested. On the other hand, new low-weight and high-sensitive portable gamma camera are also qualified on real application cases (high dose rate). 4. CONCLUSIONS Nuclear instrumentation is still in progress and the evolution of the technologies and software allow us to get imaging information in real time. The gamma camera is going to evolve and to become more compact and more sensitive. New portable gamma cameras have to be tested under high level dose rate. The alpha camera is also evolving : from the first propotype tested on site to solar blind alpha camera (to reduce the influence of light) to the new dual alpha/gamma prototype. REFERENCES [1] C. Le Goaller, G. Imbard et al. « The development and improvement of the Aladin gamma camera to localise gamma activity in nuclear installations », European Commission, Nuclear Science and Technology EUR 18230 EN, 1998 [2] O. P. Ivanov, “Control and image decoding software for portable gamma-ray imaging system with coded aperture”, IEEE NSS-MIC, conference record, Seattle (WA), October 26-28 1999 [3] C. Mahé, C. Le Goaller, F. Lamadie, Ph. Girones, F. Delmas, & al. “Imaging systems: new techniques for decommissioning”, ANS DD&R, 2005 [4] C. LEGOALLER et AL., “Gamma imaging: Recent achievements and on-going developments”, ENC 2005 [5] F. LAMADIE et AL., “Alpha imaging: first results and prospects”, IEEE 2004, Nuclear Science Symposium [6] MAHE, C., "Alpha Camera : Localization of Alpha Contamination into Nuclear Plants", IWORID, 2013

Speaker: Mr CHARLY MAHE (CEA)

15:25 NDF Strategic Plan for Decommissioning of the Fukushima Daiichi Nuclear Power Station

NDF was set up in August 2014, as an organization specializing in formulation of strategies and provision of technical support with the objectives of safe and steady decommissioning of damaged reactors. Since then efforts have been made to study the specific strategies to address major challenges of decommissioning from the mid-and-long-term viewpoint, and NDF formulated the first version of the “Strategic Plan” in April 2015, and is going to revise and formulate its 2016 version by summer of 2016.　The goals of this “Strategic Plan” are to provide a firm technical basis for the government’s mid-and-long-term Roadmap. Basic concept of the Strategic Plan The basic concept of the Strategic Plan is to continuously and promptly reduce the risks associated with the radioactive materials in the Fukushima Daiichi NPS, and risk reduction strategy is formulated for the risks which are represented by the significant effect (hazard potential) and the likelihood of loss of containment function due to radioactive materials (risk sources) such as fuels, contaminated water and waste. The major risk sources are categorized into three levels depending on the order of priority. This Strategic Plan focuses on the areas of study, the fuel debris retrieval which requires thorough preparations and has a number of challenging issues, and the waste management that requires to be addressed on a long-term basis. The technical studies on the fuel debris retrieval and the waste management are set out based on the following Five Guiding Principles to risk reduction; 1) Safe- Reduction of risks posed by radioactive materials 2) Proven- Highly reliable and flexible technologies, 3) Efficient- Effective utilization of resources (human, physical, financial, space, etc.), 4) Timely- Awareness of time axis, and 5) Field-oriented- Thorough application of Three Actuals (the actual place, the actual parts and the actual situation). Strategic plan for fuel debris retrieval (Approaches to the study on the fuel debris retrieval) Although the fuel debris is in a certain stable condition at present, it needs to be retrieved as soon as reasonably achievable by careful preparations and proven technologies, and store it in a stable condition in the site to reduce further risks. This should proceed with the following steps: 1) maintaining and management of the fuel debris in stable condition until it is retrieved; 2) safe retrieval of the fuel debris; and 3) storage of the retrieved fuel debris in a stable condition after being collected and transported. Especially among these steps, (2) safe fuel debris retrieval requires to be evaluated based on the following major issues of “identification of the location, amount, properties of fuel debris and FP distributions,” “ensuring the safety during the fuel debris retrieval work” and “the fuel debris retrieval method.” The studies on “ensuring the safety during the fuel debris retrieval work,” and “the fuel debris retrieval method” correspond to the technical requirements for the fuel debris retrieval method and consist of the following nine items: 1) securing the structural integrity of the PCV and the R/B, 2) criticality control, 3) maintaining the cooling function, 4) establishment of the containment function, 5) reduction of exposure to the workers during operation, 6) development of fuel debris retrieval equipment and device, 7) establishment of access routes to the fuel debris, 8) establishment of the system equipment and working areas, and 9) ensuring the work safety. (Options of the retrieval method and plan for selecting the scenario in accordance with the options) This strategic plan describes possible options for the fuel debris retrieval method based on the water level of the PCV which can be filled with water and the direction of accessing the fuel debris for each of the Submersion method and the Partial submersion method. After the selection of the methods to be focused on, the current status and the future actions for the nine technical requirements for the Submersion and Partial submersion methods are discussed. In addition, this plan proposes several scenarios with combinations of different methods and the plan for selecting the scenario in accordance with the conditions of each unit. The scenario of application to the actual plant is to be selected in stages in accordance with the technical development which will be a key to the success in the fuel debris retrieval method, improvement of estimation accuracy of plant conditions such as the locations and distributions of the fuel debris in each unit. The investigation required to determine the application scenario for each unit and review of the R&D plans are carried out as necessary. Reference [1] Nuclear Damage Compensation and Decommissioning Facilitation Corporation, Technical Strategic Plan 2015 for Decommissioning of the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company - Towards Amendment of the Mid-and-Long-Term Roadmap in 2015- (2015)

Speaker: Mr Kentaro Funaki (Nuclear Damage Compensation and Decommissioning Facilitation Corporation)

15:30 DEINVENTORY AND DEACTIVATION OF NUCLEAR FACILITIES AT THE ARGONNE NATIONAL LABORATORY - ALPHA GAMMA HOT CELL AND 205 K-WING FACILITIES

DEINVENTORY AND DEACTIVATION OF NUCLEAR FACILITIES AT THE ARGONNE NATIONAL LABORATORY - ALPHA GAMMA HOT CELL AND 205 K-WING FACILITIES Pancake, Daniel, Nuclear Facilities Deactivation Projects Manager, Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 E-mail address: dpancake@anl.gov Abstract: In 2006, Argonne began working on increasing the compliance posture and risk reduction efforts for ten nuclear facilities. One of those facilities, Building 212 Alpha Gamma Hot Cell Facility (AGHCF) was identified by DOE EM as the highest-ranked risk among all Office of Science (SC) facilities (Reference 1). At the time of this writing and since the 2006 period, Argonne has reduced the number of its nuclear facilities to four (including the Transportation Nuclear Facility). One of the largest contributing factors to the success of Argonne’s de-inventory program is the successful disposition of legacy materials and waste from the AGHCF and the Building 205 K-Wing hot cell facility (205K). In 2015, the Deactivation Projects and Nuclear Footprint Reduction Program have removed nearly 90% of the total amount of nuclear material from the Argonne site. This paper will focus on the deactivation program development and successes from the AGHCF and 205K deactivation processes. The Remote Handled (RH) TRU Program developed at Argonne includes several “Fist of a Kind” waste streams that allowed disposition of Fuel Examination Waste (FEW) and Separations Science waste from the hot cells. These Argonne Programs are already being implemented at other DOE TRU Waste sites, and have the potential to add significant value to those Sites’ Programs. The inventory reduction from the AGHCF alone amounts to approximately 66x the HazCat 2 (HC-2) SOF TQ. These risk reduction and compliance efforts included removal and disposition of nuclear materials from two additional HC-2 Nuclear Facilities (hot cells), that have since been de-inventoried and downgraded to radiological facilities. These Projects were accomplished on-time and under budget, while transitioning from “vintage” Safety Bases to 3009/3011 compliant Safety Bases (Basis for Interim Operation [or BIO] for Deactivation work). As a result of these de-inventory efforts, Argonne has dispositioned more curies of RH TRU waste to the Waste Isolation Pilot Plant (WIPP) than all other DOE Lab Programs combined (through close of FY15). This successful completion of the projects advanced de-inventory efforts at the Laboratory by several years. The deinventory process successfully dispositioned approximately 10,000 irradiated test specimens from the AGHCF. 1. INTRODUCTION Prior to the establishment of the Deactivation Projects, DOE EM identified the Building 212 AGHCF as the highest-ranked risk of all legacy DOE-Office of Science (SC) nuclear facilities in 2008. Since then, Argonne has maintained an aggressive path forward in removing inventory from the AGHCF, and by the end of FY15 had reduced the AGHCF inventory to .024x HC-2 SOF. The inventory of the AGHCF at the end of FY15 contains only the surface contamination values, as all of the source term has been removed. 2. METHDOS Argonne’s NFRDP and ARRA efforts led directly to the establishment of several new waste streams within the RH TRU Program. Fuel Examination Waste (FEW) and Solidified Separations Science Waste (S3W) streams were developed to drive disposition of irradiated fuel test specimens, SWARF, and solidified liquids generated from UREX experiments in various hot cells. The approval of these waste streams allowed the disposition of 100% of the 205 K-Wing RH TRU, and approximately 80% of the AGHCF discrete inventory, to WIPP near Carlsbad, NM. This leading effort promises complex-wide benefits in the future. 3. RESULTS • De-inventory and deactivation of Building 205 K-Wing Hot Cells (HC2 to Radiological Facility (RF)) • De-inventory and Deactivation of Building 200 MA/MB Wing multi-story Hot Cell Facility (HC2 to RF) • De-inventory of the AGHCF (Spent fuel repatriated to INL, over 500 drums of RH TRU shipped to WIPP, reduction of approximately 66X HC2 SOF) • Development of 5 new RH TRU characterization methods, and approval of 5 Tier-One Change Requests through the US EPA and CBFO • AGHCF won the UChicago-LLC Award for Team Safety Performance in 2012 and 2013 4. CONCLUSIONS Argonne’s important future mission work will benefit greatly from the success of the NFRDP, the accelerated pace of those successes provided by the ARRA funding, and by maintaining a record for safety and production that demonstrates the Laboratory’s expertise in executing high-risk work on a daily basis. Argonne maintains the core set of expertise in the realm of RH TRU FEW Program Development and Management, and has been instrumental in partnering with other DOE facilities assisting in implementing similar programs and project work.

Speakers: Ms Cindy Rock (Argonne National Laboratory), Mr Daniel Pancake (Argonne National Laboratory)

15:35 KRR-1 Decommissioning

The first research reactor in Korea, KRR-1, is a TRIGA Mark-II type with open pool and fixed core (Figure 1). Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The aim of the decommissioning activities is to decommission the reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR-1 reactor was decided to be preserve as a historical monument. KRR-1 reactor for removing all radioactive materials and the bio-shielding concrete and building structure were remain for the remodeling for monument. Dismantling of the KRR-1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan changing, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), and conditioning the radioactive waste for final disposal, and final statuses of the survey and remediation for free release of the site and building.

Speaker: Mr Seungkook Park (KAERI)

15:40 Immediate Dismantling of a Large Fleet of LWR NPPs: Consequences for Spent Fuel and Waste Management

The experience feedback shows that the complete dismantling of one Light Water Reactor (LWR) unit is an activity now globally well under control (reactor pressure vessel and its internals removal and cutting included). However some specific industrial difficulties may arise in view of the dismantling of numerous LWR units simultaneously. Indeed more than 300 LWR units are currently operated worldwide, which have been commissioned these last 50 years. So in some countries, many units may be permanently shut-downed over a period of few years in the next decades. Such a situation addresses the issue of the overall management of large quantities of spent nuclear fuel (SNF) and decommissioning waste for the concerned units. For example, in France, the legislative and regulatory framework for the nuclear facilities favors their dismantling “as soon as possible” after their permanent shut-down which implies as well to limit the duration of the transition period from operation to decommissioning. Furthermore 58 LWR units have been commissioned between 1977 and 1999 in France – on average more than 2 units per year. In this context, the operator (Electricité de France – EDF) plans to remove quickly the SNF then to perform dismantling actions immediately after the permanent shut-down of the LWR units. One issue is to remove the SNF from all the relevant units, even if this removal is simultaneous in many units (permanently shut-downed or still under operation). Similar issue has to be taken into account regarding the radioactive waste (RW) produced by the dismantling and clean-up actions, notably the RW that cannot be disposed of in a near surface repository. One method followed by the French technical support organization (Institut de Radioprotection et de Sûreté Nucléaire – IRSN) to review these issues is the use of estimates of flows of SNF and RW, based notably on a phasing-scenario and a planning template defined for the decommissioning of the units of one nuclear power plant (NPP) and coupled to an overall schedule for phase out all the units of the fleet. These estimates relative to the next decades can be compared to the current experience feedback of flows of SNF and RW for units under operation, in order to identify risks when facing decommissioning. The risks highlighting are driven by key parameters (as duration of the main dismantling actions) of the estimates which can be adapted to minimize their impact. On this basis, it is possible to identify the key-factors to dismantle each unit of NPPs and phase out the fleet regarding SNF and RW management. It is noteworthy that this work needs to be done in any case upstream the studies and the implementation of dismantling actions. Finally it can be underlined that another issue is the human resources (staff, skills and knowledge) necessary to perform all the decommissioning actions, but this aspect is not addressed here.

Speaker: Mr Denis DEPAUW (IRSN)

15:45 Decommissioning and Demolishing Stacks from legacy fuel cycle facilities at Sellafield Site

The Sellafield skyline provides visual markers on the progress being made in decommissioning. The most notable change in the last few years has actually been one of construction of a new 123m high stack which is designed to replace current assets and also provide an expanded radioactive ventilation service for at least 100 years, supporting the decommissioning and demolition legacy buildings. The planning for the decommissioning and demolition of two of the most iconic structures at Sellafield - the First Generation Reprocessing Stack and Pile Chimney – are well advanced. Although both stacks fundamentally provide(d) the same functionality (discharge process steam and heat together with dilution and dispersion of radioactive aerial effluent due to their height) they present very different challenges. However, the overall risks are similar – damage to adjacent facilities. First Generation Reprocessing Stack This 61m operational stack is used to dilute and disperse 80% of the site airborne radioactive material discharges. It is situated on the roof of the now redundant first generation separation plant facility which itself is located within a cluster of operational Magnox reprocessing and waste management facilities at the heart of the Sellafield Site. Due to its age it does not meet current seismic resistance standards. Removal of the stack therefore addresses the safety risk associated with design basis earthquake events. Following detailed engineering and safety assessments work has already commenced with the installation of a temporary external passenger and goods lift to the side of the building and a further lift will be attached to the stack. A mock stack has been constructed off-site to trial the self-climbing platform – this provides the opportunity to train operators and fine-tune operational practices. The method of demolition is tried and tested. Hydraulic jaws will size reduce the concrete and the internal metal flue will be cut using plasma arc. It is expected that the demolition will reduce the size of the stack by around a metre a week. To put the overall scale of the task into perspective the amount of waste has been estimated as 600 tonnes of concrete and rebar plus 25 tonnes of stainless steel fuel liner. Pile Chimney The adjacent facility and stack was the site and high level discharge point for the UK’s worst radiological event – the 1957 Windscale fire. Notwithstanding the additional radiological considerations related to the pile chimney as opposed to the first generation reprocessing stack, the location of the pile chimney provides a different demolition challenge and associated opportunity. As with the stack, detailed assessments are required to support safe construction and use of the tower crane and associated equipment. The solution for the pile chimney is to use a luffing tower crane. More commonly seen being used in high-rise city development areas they requiring little room for slewing. At a total height at full extension of 151m the crane will be tied-in to the pile chimney at a number of locations. The lifting capacity of 12 to 14 tonnes it will be used to remove the concrete blocks and supporting the cutting of over 1.2km of wire.

Speaker: Mr Steve Slater (UK)

15:50 Use of highly selective ion exchangers in different decommissioning projects

Fortum's highly selective ion exchanger media CsTreat®, CoTreat®, and SrTreat® have been used in several nuclear waste treatment projects. The media has found use at operating power plants, decommissioning sites, and for example also at reprocessing facilities. Most notable recent example of the use of the Treat products is the Fukushima accident site's water treatment program, where CsTreat and SrTreat have been used. In this paper two examples of the use of the Treat products in recent decommissioning projects (Doundrey and Bradwell in the UK) is presented. Further, the role of the Treat products in the future decommissioning of Loviisa plant in Finland is discussed. Finally, the nuclide and boron removal system is at Paks in Hungary is presented. 1. Doundreay site, UK: Na and Na/K coolant of fast breeder reactors were dissolved and the resulting water containing salt ions was decontaminated using CsTreat. Total of 57 tons of Na/K coolant from the Doundreay Fast Reactor (DFR) and 1500 tons of Na coolant from the Prototype Fast Reactor (PFR) were purified during years 2002-2012. Cs-134 and Cs-137 activity in the purified liquid was below detection limit, with the highest achieved decontamination factors for Cs being as high as 4 million. 2. Bradwell NPP, UK: Fuel element debris (FED) containing magnesium-aluminum alloy (Magnox) material is treated in the Discharge Abatement Project (ADAP). FED is recovered from underground storage silos, dissolved, and the resulting dissolved salt solution is treated using CsTreat and CoTreat in order to remove both Cs and activated corrosion products. Very high selectivity towards the target nuclides is important, as the solution contains very high background of competing Mg ions. Operation started in 2013 and very first results are becoming available. 3. Loviisa NPP, Finland: Loviisa power plant is two unit VVER-type plant. Current operation licenses are expiring in 2027 and 2030. It is planned that during the decommissioning stage CsTreat and CoTreat will be used to purify most of the radioactive water at the plant. Waters include the primary circuit, fuel pools, and different tanks (including boric acid storage tanks) so that the waters can be released to the sea. Only some dozens of litres of secondary waste, i.e. spent ion exchanger mass, will be produced. This provides considerable cost advantage compared to solidification of several hundred cubic meters of boric acid water. Similar approach of nuclide removal can be used at different PWR plants located on the seaside, both during operation and decommissioning. 4. Paks NPP, Hungary: Paks is currently operating NPP with four VVER-type units. At Paks Fortum's CsTreat material is used to remove Cs from the evaporator concentrate. Hungarian technology is used to remove corrosion products. Additionally, a large fraction of boron is removed (in the form of sodium tetraborate) using Fortum's technology, as only small amounts of boron can be released to the river Danube. After treatment the concentrate is nearly free of radioisotopes and contains only a fraction of original boron concentrate and thus can be released to the river Danube. Recovered solid boron (sodium tetraborate) can be released from the radiological control. A process which combines nuclide and boron removal can be used to treat boric acid waters when little or no boron release is allowed. This process in relevant both during operation and decommissioning.

Speaker: Dr Jussi-Matti Mäki (Fortum Power and Hest Oy)

15:55 Benefits from R&D For Decommissioning and Remediation projects

CEA (French Alternative Energies and Atomic Energy Commission) is both the operator of important nuclear facilities all over the nuclear cycle, in charge of major new built or Decommissioning and Dismantling (D&D) projects and a R&D group with dynamic policy of technology transfer. The position of CEA in D&D is unique because of the number and the wide diversity of facilities under decommissioning, with some high level of contamination. Innovative solutions are being developed in 6 main axes to protect the operators, to minimize the overall costs and the volumes of waste, especially used when preparing D&D operations: Investigations in the facilities, Radiological measurement of waste, Technologies for hostile environment, Decontamination of soils and structures, Waste treatment and conditioning and Methods and Information Technology (IT) Tools for project and waste management. The last developments are shown and examples of industrial applications given. CEA is willing to share actions in partnership with other operators or with industrials dealing with the same problems to solve.

Speaker: Mrs Christine GEORGES (CEA)

16:00 RISK MANAGEMENT FOR DECOMMISSIONING OF FACILITIES - THE DRIMA APPROACH

In order to provide assistance and feedback on the application of risk management methodologies in preparation of and during decommissioning, IAEA’s International Decommissioning Network (IDN) launched in December 2012 the “Decommissioning Risk Management Project – DRiMa Project” which was completed in November 2015. The DRiMa project developed an approach for risk management which inter alia distinguishes between the need to manage uncertainties associated with assumptions (related to initial decommissioning plans) or with strategic decisions (setting the frame for final decommissioning plans), and the need to manage (operational) risks during the implementation of decommissioning activities (project risks). Central elements recommended by the DRiMa Project are the use of prompters to systematically analyze assumptions, strategic decisions and their uncertainties and (operational) risks and the use of registers to systematically manage these entities.

Speaker: Dr Joerg Kaulard (TUEV Rheinland Industrie Service GmbH)

15:00 → 16:30 Session 4B - 3: Technical and Technological Aspects of Implementing Environnemental Remediation Programmes - Parallel Session

15:00 Lesson Learned from the Implementation of Decontamination iin Japan

Lesson Learned from the Implementation of Decontamination in Japan Hiroyuki　Kuroda International Cooperation Office for Decontamination Radioactive Materials, Environment Management Bureau Ministry of the Environment(MOE) ,Japan Upon release of radioactive materials by the accident at Fukushima Daiichi Nuclear Power Station on March 11, 2011, the Government of Japan as well as the prefectural and the municipal governments have been taking measures to decontaminate polluted soils in order to reduce the impact of radioactive pollution of the environment on human health and the living environment as soon as possible. In the efforts for the decontamination, all available resources including those from the central and the local government offices, research institutions, and private cleaning operators are put together, along with the scientific and technical knowledge available from Japan and abroad. Our primary objective is to eliminate the recurrence of such a disaster in the future, but in the meantime, disclosing and sharing our knowledge, experience, and lessons obtained through the decontamination efforts this time with domestic peers and the international community will be significant to accelerate the decontamination work in Japan and minimize the potential damage in future accidents in Japan and abroad for the implementation of expeditious and efficient decontamination. Therefore, the Ministry of the Environment(MOE) published “Decontamination Report” and the report comprehensively compiles the basic policy of the decontamination and implementation framework, knowledge about the management of decontamination projects based on the actual decontamination operations on-site, together with the procedure, conditions and effects of individual decontamination techniques, by mainly focusing on the decontamination operations performed by MOE. As the report shows, the basic idea of decontamination works of the national and the local governments toward residents are as follows; ・Principal radionuclide of environmental pollution is cesium. ・Decontamination should be conducted to reduce the impact on human health or the living environment not only in housing areas and public facilities, but also in more diverse and wide areas such as roads, farmland, forests (living zone only), and the like. ・Decontamination should be implemented as soon as possible in order to realize return of residents in the evacuation areas, to secure their safety, and to rebuild their lives as soon as possible. ・Residents' opinions and familiar way of life should be respected; decontamination should be conducted by paying due consideration to the protection of private rights and maintenance of the community. Based on those principles MOE, other concerned ministries, the Prefectural Government of Fukushima, and other local governments strived to implement decontamination of not only residential areas and public facilities but also of diverse and wide areas of land including roads, farmland, forests adjoining residential areas on a trial-and-error basis, keeping in mind that residents can safely and quickly return home and rebuild their lives. Implementing decontamination works, Some of the lessons we learned are as below; ・Importance of basic technology for decontamination ・Well-organized management for decontamination projects ・Securing the quantity and quality of decontamination workers ・Keeping well-communications with local residents MOE will promote more effective and efficient decontamination works in and around Fukushima Prefecture and share both domestically and internationally our experiences, lessons and knowledge learned through the environmental remediation efforts. REFERENCE 1. Ministry of the Environment ,Japan(MOEJ),2015 “FY2014 Decontamination Report-Digest Version-“ Page1-20, 75-78 2. Ministry of the Environment ,Japan(MOEJ),2015 “Progress on Off-site Clean-Up and Interim Storage in Japan” Page 1-38, 53-57 3. Environmental Information Science(Journal in Japanese) No.2 ,Volume 44, 2015 Hiroshi ONO, “Progress and Challenges on Off-site Decontamination based on the Act of on Special Measures concerning the Handling of Radioactive Pollution” Page 27-32 .

Speaker: Mr Hiroyuki KURODA (Ministry of the Environment,Japan)

15:25 The Current State of Food Regulations in Fukushima: Environmental Consequences of the Nuclear Accident

The Fukushima Daiichi nuclear disaster caused radioactive contamination in a wide range of Fukushima Prefecture. Because 70% of the land in Fukushima is forest, analyzing the situation in the forests will provide us with an estimation of the conditions of the environment in Fukushima as a whole. Decontamination of the forests is not being carried out, and only 1% of cesium in the forests is estimated to be washed out by rain annually, which gives us a reason to believe that most of the cesium still remains in the forests. In fact, cesium has been detected in animals inhabiting the forests since they have stayed in the contaminated circumstances. Right after the disaster, the Fukushima government started monitoring radioactive substances in the muscles of wild animals in the forests which were used for food. This effort was made to provide information to secure the safety of hunters as well as to remove the anxieties of the citizens. Additionally, the Fukushima Prefectural Centre for Environmental Creation has been examining the change in the amount of cesium in the muscles of wild animals. A paper published in Germany reports the time series variation of Cs137 concentration in the muscle of roe deer that were living in the forest in Bodenmais for 20 years after the Chernobyl accident. We compared their results with the annual changes in the amount of cesium found in Japanese deer living in the Aizu mountain area (in western Fukushima Prefecture). We believe the comparison is meaningful in that it provides a criterion to compare the environmental contamination of the two regions. The forest in Bodenmais is about 1500 km away from the Chernobyl Nuclear Power Plant, but right after the accident, roe deer muscle had a concentration of over 10,000 Bq/kg. 20 years later, however, the concentration had decreased to 1,000 Bq/kg. The Aizu region, on the other hand, is 100 km away from the Fukushima Daiichi Nuclear Power Plant. Right after the accident, there were some Japanese deer that had a concentration of about 300 Bq/kg in their muscle, but now (4 years later) no Japanese deer has been found with a cesium concentration over 100 Bq/kg. Based on these data, it is conceivable that the radioactive contamination level in Fukushima is much lower than that in Europe. Four years have passed since the nuclear accident. Currently, the state of food regulations in Fukushima is as follows. Fukushima Prefecture produces various kinds of agricultural crops, including rice, and fruit such as peaches. Immediately after the accident, since rice is a staple of the Japanese diet, the prefecture decided that all rice should be examined for radioactivity. Every year, ten million bags of rice (each bag containing 30 kg) are harvested and sold on the market. According to the results of the examination, only 71 out of the ten million bags were found to be contaminated with 100 Bq/kg of cesium or more in 2012, 28 bags in 2013, and two bags in 2014. We believe that we decreased the amount of cesium in food on the market to a great degree by such measures as spreading potassium during the growing period of rice to inhibit it from absorbing cesium. For the first time after the disaster in November, 2015, the EU revised the regulations of food imports from the North-Eastern region of Japan. However, the import of some agricultural products, such as rice, is still prohibited. Fukushima Prefecture has been actively conceiving ways to improve the safety of agricultural crops on the market and has been observing the situation. As a result, the food sold on the market adequately meets international safety standards. In the presentation, the current state will be reported based on the data.

Speaker: Mr Shigeaki TSUNOYAMA (Director)

15:50 DEVELOPMENT AND DEMONSTRATION OF AUTOMATED PRETREATMENT SYSTEM AT INTERIM STORAGE FACILITY FOR CONTAMINATED MATERIAL GENERATED FROM DECONTAMINATION OF OFF-SITE AREAS SURROUNDING THE FUKUSHIMA NPP

Contaminated material (soil/waste) generated from the decontamination of large contaminated areas surrounding the Fukushima Daiichi Nuclear Power Plant is being collected and stored at temporary storage facilities at all municipalities in Fukushima prefecture. The contaminated material will be transported to centrally-located Interim Storage Facilities (ISF) where it will be sorted, treated, and then placed for interim storage for up to 30 years. The current estimated volume of material that will be processed at the ISFs is 28 million m3 and effective technologies to help volume reduction are expected. Following acceptance inspection at ISF, contaminated material will be pretreated, i.e. segregated with regard to combustibility and radioactivity concentration. Such pretreatment work with enormous volume material is a key for early and safe implementation of interim storage. Obayashi Corporation has developed a series of automated pretreatment technologies which contributes to prompt operation, worker’s safety and volume reduction.

Speaker: Ms Shoko YASHIO (Obayashi Corporation)

16:15 Lessons Learned in Fukushima Environmental Remediation in the standpoint between Governmental Bodies and Local Residents - from viewpoint of one engineer involved –

Lessons Learned in Fukushima Environmental Remediation in the standpoint between Governmental Bodies and Local Residents - from viewpoint of one engineer involved – Tadashi INOUE Research Advisor to CRIEPI, Emeritus The Accident of the Fukushima Dai-ichi Nuclear Power Plant posed a wide range of environmental contamination by mainly radioactive cesium-134, and -137. After 5 years passed, hundred thousands of residents are still evacuated, though the remediation activity has been progressing. Before the accident, Japan never expected such catastrophic disaster and never prepared for remediation and storage of a large volume of radioactive materials. Up to the present, effective lessons learned are summarized in the IAEA review mission reports, IEM-4 Conference report, and so on by studying the remediation activities in Fukushima. This paper describes supplemental points of lessons learned, which are drove from the activities in the academic society and in the role of advisor to local governments. Lesson learned 1: Responsible parties to manage/govern all activities for remediation including monitoring should be required. At primitive stage of remediation just after the accident, wide range of players, such as scientists, engineers, NPOs, etc., engaged in the remediation individually, which caused to uniformity on remediation and storage of contaminated material. In addition, radiation dose and radioactivity of soils were measured independently, which caused a feeling of insecurity on local residents. After this stage, Japanese government has taken these roles, that is, integration of the remediation works and integration of monitoring data. Guideline published by MOE is an effective for making a uniform remediation and waste management. Lessons learned 2: Effective decontamination means should be selected Even if the same remediation means could be applied, the decontamination factor changes within a wide range depending on the characterization of the contamination. A hard remediation method, such as top-soil removal, produces a large amount of radioactive materials to be stored and disposed of. Decontamination method should be selected based on the cost and benefit with the acceptance of local residents. Lesson learned 3: Effective application of reference level should be required The ICRP recommends applying on a reference level in the range of 1 – 20 mSv as an annual effective dose. The IAEA safety guide suggests that “if the projected annual effective dose for the representative person falls between 1 and 20 mSv/y the benefits of remediation should be justified and a reference level established within that range”. In case of the Fukushima accident, the target of the remediation is determined with 1 mSv/y additionally for a long-term goal. A certain number between 1 – 20 mSv/y should be applied in the wide range contamination based on the cost and benefit. In the meanwhile, we have to notice the public has a doubt why the number is different in a short-term and a long-term goal, if the certain number in 1 – 20 mSv/y is a safe number. Lesson learned 4: Priority of the remediation is a strategic issue and should be determined along the real situation - Most high priority area for remediation should be an area where residents are living and, then, an area where evacuee will return after the remediation - High contamination area may not be a target of the first remediation activity due to the fact radiation dose does not decrease enough after one-step remediation, that means the mid- and long term activity for homing. - Public building, such as schools, municipality office, etc. should be a high priority. Lesson learned 5: Acceptance of community and property owner is essential for the progress of remediation - Provided by receiving the acceptance from local community/municipality, PTA, etc. smoothly progressed is the remediation of public building, sport field, park and so on - The acceptance of private owners for houses, gardens, is crucial and critical, and consumes a time. Lesson learned 6: Role of academia Local residents feel uneasy what the governmental bodies said, and have various voices for requirements. In addition, they are worried which opinions they should trust among various ones. Therefore, the academia should play a role to transfer the correct knowledge based on the authorized scientific evidence with comprehensive manner, and to take an interface between governmental bodies and residents. At last, the followings also come up for the issues to be concerned, - Different voices by scientists and engineers should be avoided due to making confusion and distrust for residents. - Spread of noise and rumor should be avoided. Some of peoples living outside Fukushima give negative influence by spreading bad rumor. References 1. The follow-up IAEA International Mission on remediation of large contaminated areas off-site the FDNPP, 14-21 October 2013. 2. IAEA Report on Decommissioning and Remediation after a Nuclear Accident, Int. Exp. Mtg, 28 Jan. – 1 Feb. 2013, Vienna, Austria.

Speaker: Dr TADASHI INOUE (Central Research Institute of Electric Power Industry)

16:30 → 17:00 Coffee Break

17:00 → 18:00 Session 4A - 4: Technical and Technological Aspects of Implementing Decommissioning Programmes - Parallel Session

The purpose of this session is:
• To review progress in decommissioning technologies and cost estimation over the past decade and to identify challenges and needs in the future
• To review current developments, including from damaged facilities, with a view to identify existing gaps in knowledge and needed improvements

17:00 → 18:00 Session 4B - 4: Technical and Technological Aspects of Implementing Environmental Remediation Programmes - Parallel Session

The purpose of this session is:
• To review the available technologies for environmental remediation
• To define the existing gaps in knowledge and needed improvements with the aim of facilitating the implementation these activities.

Thursday, 26 May

09:00 → 11:00 Session 5A - 1: Optimizing Waste and Materials Management in Decommissioning - Parallel Session

The purpose of this session is:
• To discuss the developments and challenges being faced dealing with materials resulting from decommissioning activities
• To consider the integration of decommissioning and waste management activities from initial planning to final waste disposal (including management of problematic materials)

09:10 OVERALL OPTIMISATION OF DISMANTLING RADIOACTIVE WASTE MANAGEMENT IN FRANCE

It is estimated that by 2030, the dismantling of nuclear installations in France will have generated more than 700,000 cubic metres of radioactive waste. As an agency tasked with offering disposal solutions for all radioactive waste, Andra has an important role to play, whether in adapting waste management conditions to the specific needs of dismantling, studying ways of recycling some radioactive waste to save space in disposal facilities, or securing dismantling scenarios. This approach requires close collaboration with other stakeholders involved in dismantling (such as licensees, operators and transporters), in order to set up the most appropriate tools and organisation system and optimise all aspects of waste management. Regarding radioactive waste, a successful dismantling operation is one that has sought to optimise every aspect of radioactive waste management, from the production of waste to its disposal, including waste characterisation, processing, packaging and transport. Optimisation seeks to minimise personnel exposure to ionising radiation, as well as minimise costs, limit waste production, and make optimal use of capacity at waste disposal facilities. As a provider of long-term management solutions (including disposal) for all radioactive waste, and because all the stages that precede disposal depend on how this waste is managed, Andra plays a part in the dismantling process and must therefore coordinate its actions with those of the other stakeholders involved in dismantling: nuclear licensees responsible for the outcome of their installations, other operators active on the sites, industrial partners equipped with waste processing tools, and transporters.

Speakers: Mr Frédéric LEGEE (Andra), Mr Michel DUTZER (Andra)

09:35 Integration of decommissioning and RW management operations in Spain

The radioactive waste produced during decommissioning operations have to be characterized, treated, conditioned and store on site under the Decommissioning License, and they must to meet the requirements established in the: • Transport Regulations • Disposal Acceptance Criteria and Package specifications Site (El Cabril repository) Decommissioning and radioactive waste management operations in a decommissioning site are very much affected by the waste requirements in the disposal site. So, decommissioning and waste management are very much interdependent and an integrated approach is required Additionally, it is generally required by the regulator to reduce the production of radioactive waste as low as reasonably achievable, so in activity as in volume, through the application of adequate measures of design, operation practices and decommissioning, including the recycle and reuse of materials. The planning and implementation of decommissioning strategies require the support of a radiological inventory. One of first activities in a decommissioning project is the estimation of a radiological and physical inventory of the material, including soils. Once the Radiological Inventory has been estimated, and taking into account the acceptance criteria and waste packages of the repository and clearance criteria of material and site, it is obtained a classification in material’s streams and packages. According to these material streams, the waste management routes are established and, in application of the principle of volume reduction, the following waste management issues are studied: o Large components cutting o New containers for large and highly activated/contaminated components o Capacity of LILW & VLLW repositories and GTCC storage o Waste sorting in situ, according to defined streams o Waste Radiological characterization and clearance processes o Waste volume reduction techniques (mechanical cutting, more efficient packaging/conditioning, use of new containers, compaction and shredding, etc.) and decontamination techniques (chemical decontamination and blasting of metals and soil washing) Some changes in the decommissioning and waste management process could require design modifications in the decommissioning license and in the waste disposal site license, such as: • Commission new waste management facilities at the decommissioning site • Commission new disposal units in the waste disposal site The main conclusions and lessons learned that ENRESA has derived from an integrated approach of decommissioning and radioactive waste management may be summarized as follows: • Detailed characterization and management plans should be established for the different waste streams in order to ensure a fast and smooth logistic of materials. • Segregate “in situ” as much material as possible to minimize quantities of radioactive waste. • Clearance contributes to the minimization of wastes, resulting from dismantling and remediation, and allows recycling of materials. The material clearance process should be performed at industrial scale with large size containers and should be well-tested and integrated with the decommissioning operations. • Very low level waste disposal is complementary to clearance and allows disposal of wastes with activities 10 to 100 times the clearance levels. • Volume reduction/decontamination techniques play a significant role in reducing the amount of waste to be disposed and contribute to a more efficient and sustainable waste management system. • Segmentation of activated components should be performed by mechanical cutting in order to minimize secondary wastes • Conditioning/storage of highly activated/contaminated wastes will require a specific auxiliary facility. This facility replicates the conditioning process of disposal units (CE-2a) in El Cabril • Large containers should be used for highly activated/contaminated components in order to reduce segmentation costs and operator doses. The removal and management of major components in large pieces or in one piece should be considered for future projects • Reactor internals (non-LILW) need to be stored together with the spent fuel at a dry storage facility • Site remediation generates large volumes of material to be monitored and large amounts of very low level waste to be disposed.

Speaker: Mr Nieves Martin Palomo (ENRESA)

10:25 RESEARCH REACTOR BR-10 - TESTING GROUND FOR CONDITIONING OF RADIOACTIVE WASTE OF ALKALINE LIQUID-METAL COOLANTS UNDER DECOMMISSIONING PROJECT

In 2002 after successful work over a period 46 years the research reactor BR-10 (RR BR-10) was shut down and transfer into decommissioning status. The total amounts of alkaline coolant radioactive waste are approximately 18 m3 with the total activity in excess of 37 TBq (1000Ci). These are: sodium - 9 m3, including sodium from 16 cold traps oxides; alloy NaKHg - 4,5 m3 with a ratio 44% - 48,5% - 7,5% and 4,5 m3 with a ratio 95% - 5% - 0,02%. Later, treatment technologies with radioactive waste of alkaline liquid metal coolant and with polluted equipment were developed as a result of decision of scientific-technical challenges. Currently in the rector building introduce in operation the testing ground for transferring into safety stable condition radioactive waste of alkaline liquid metal coolant from RR BR-10. Decommissioning project includes units which founded on the developed treatment technologies. Namely: Magma  for solid-phase conditioning the total volume of alkaline coolant radioactive waste; Getter  for purification alloy NaK from mercury admixtures; Luisa-RW  for neutralization nondrainable residual radioactive waste of alkaline coolant. These technologies have experimental and estimated justification for industrial application under decommissioning project of RR BR-10. For conditioning the total volume of discharge alkaline liquid metal coolant in safety condition was created unit Magma. The principle of operation of the unit is based on the solid-phase oxidation technology by slag from the copper-smelting industry [1-3]. Other main challenge at the RR BR-10 is treatment with radioactive waste of alloy NaK which impure of mercury admixtures. For solution of this problem was created unit Getter, the principle of which is based on the technology of purification alloy NaK from mercury with a hot magnesium getter [2-3]. For removal nondrainable residual radioactive waste of alkaline liquid metal coolant from inside surface of individual equipment was created unit Luisa-RW. The principle of operation of the unit based on the technology of neutralization nondrainable residues by nitrous oxide with subsequent grouting of obtained salt phase directly in the volume of equipment. A priority unit Luisa-RW is handling with cold traps oxides [2-3]. Currently, at the testing ground is produced complex tests of equipment and adjustment of technological conditions. The operation of the testing ground, which is scheduled to begin in 2016, is assumed:  unit Magma  conditioning of sodium from 1 and 2 circuits (approximately 6 m3);  conditioning of sodium from cold traps oxides (approximately 3 m3);  conditioning of cleared alloy NaK from mercury (approximately 9 m3);  unit Getter - purification alloy NaKHg from mercury (approximately 9 m3);  unit Luisa-RW - neutralization nondrainable residual in 16 spent cold traps oxides and other individual equipment.

Speaker: Mr Vladimir SMYKOV (Institute for Physics and Power Engineering, Obninsk, Russia)

09:00 → 13:30 Session 5A - Poster

09:00 Alternative ILW Management Approach for Magnox Swarf Storage Silos at Sellafield

The Magnox Swarf Storage Silos (MSSS) has been used to hold intermediate level waste primarily from Magnox fuel decanning, for decades. The aging facility poses an intolerable environmental hazard and the waste within needs to be retrieved to allow decommissioning of the facility. The previous strategic approach led to complex engineering solutions and a perception of potentially significant consequences from hydrogen or ignition events from the presence of reactive materials within the waste. The main issues affecting retrieval of the waste are centred on limited understanding of both the current composition of the waste and the behaviours likely to be exhibited during retrievals and processing. The two main risks considered within the reactive materials work were: • Generation and evolution of hydrogen gas • Presence of self-igniting (pyrophoric) materials leading to waste fire A combination of broad research and empirical experiments has led to improved understanding of the corrosion processes in MSSS. A more accurate picture of the waste inventory at retrieval has been developed, as well as a more realistic basis for waste behaviour during retrieval and downstream processing. The research and development focused initially on the corrosion mechanisms of uranium and magnesium; this was used to model the waste tipping, storage and planned retrievals. The predicted waste container payloads are now expected to present a significantly lower hydrogen and pyrophoric challenge than the previously assumed basis. Investigations into the reactivity of the more hazardous components, such as uranium hydride, which might cause problems for safety and engineering followed. Technical specialists carried out assessments on the formation, survival and reactivity of these reactive materials in order to evaluate the pessimisms constraining the design process, and to identify opportunities to reduce these restrictions. The results were so favourable that not only was the objective of underpinning the assumed design basis achieved, many additional opportunities were identified. An early opportunity realised was a change from a sealed inerted flask into a passively vented flask. However, the most fundamental opportunity is for accelerated MSSS risk and hazard reduction to include the storage of waste in an interim store. The strategic review identified a much simpler approach to retrieval and processing of the MSSS waste involving interim storage of the waste in its raw form in a double-skinned steel box with an intermediate grout bund, which would then be placed in one of the Sellafield site product stores until an appropriate final conditioning plant is ready, prior to ultimate disposal in Geological Disposal Facility.

Speaker: Mr Alan Parry (UK)

09:00 APPLICATION OF DECOMMISSIONING COSTING FOR RESEARCH REACTORS USING CERREX CODE

APPLICATION OF DECOMMISSIONING COSTING FOR RESEARCH REACTORS USING CERREX CODE KRISTOFOVA, Kristina1; PARK, Seungkook2; ZACHAR, Matej3; DANISKA, Vladimir3 1 KIND Consultancy, Zvoncin 299, 91901 Sucha nad Parnou, Slovakia 2 Korea Atomic Energy Research Institute (KAERI), 989-111 Daedeok-daero, Yuseong-gu, Daejeon, 305-353, Korea 3 DECOM, a.s., Sibirska 1, 917 01 Trnava, Slovakia E-mail address: kristina.kristofova@kind-consultancy.com Abstract: The main purpose of the paper is to present the possibility for application of recommended principles and methodology for decommissioning costing of research reactors by using CERREX (Cost Estimation for Research Reactors in Excel) code. For that purposes the model calculation case was developed using the inventory data from Korea Research Reactor 2 – KRR2 (TRIGA Mark III type) [4] and pre-defined set of calculation data implemented to CERREX. To create an interface between the inventory database and CERREX calculation module, the special Excel template was developed taking into account physical (mass, surface) as well as radiological parameters (contamination, dose rates, nuclide vectors, limits and conditions). Finally, the calculated results followed by the results of sensitivity analysis are discussed. 1. INTRODUCTION The decommissioning of non-power-generating facilities such as research reactors, owing to their worldwide distribution, is of special importance within the IAEA activities. To support the decommissioning cost estimation for research reactors as a part of decommissioning planning process, the calculation code CERREX [2] was developed within the IAEA projects. CERREX code is developed in line with the IAEA recommendations for decommissioning costing [3] and fully implements the ISDC (International Structure for Decommissioning Costing of Nuclear Installations) structure and costing methodology [1]. 2. METHODS In order to assemble CERREX decommissioning costing case and perform sensitivity analyses it was necessary at first to collect inventory data on KRR2 reactor. For that purpose an inventory database template in Excel was used including building–floor–room structure data of the facility and detailed equipment inventory parameters. Equipment parameters represented the following type of data: 1. Identification data – equipment designation, allocation within the facility structure and technological system, relevant ISDC No. 2. Physical inventory data – mass, surfaces, CERREX category of equipment and dominant material 3. Hazardous inventory data – hazardous material and possible hazardous waste 4. Radiological inventory data – internal/external contamination, activation, dose rate and corresponding radionuclide vectors and reference dates 5. Calculation data – recalculation of radioactivities in time and determination of resulting waste streams At second, the further steps regarding development of model KRR2 decommissioning costing case in the CERREX code should be briefly summarized as follows: 1. Implementation of the inventory database to the CERREX code i.e. definition of relevant partitioning coefficients for defined waste categories; 2. Definition of input parameters (duration, composition of the workgroup, expenses or investments) for period dependent activities; 3. Definition of calculation parameters (e.g. labour rates, manpower unit factors and cost unit factors, work difficulty factors) respecting the ISDC methodology; 4. Analysis of the obtained results – basic calculation case; 5. Performance of sensitivity analysis to estimate the impact of changing input parameters (e.g.: decommissioning start date, level of contamination, scope of the project, labour costs) on calculated results. 3. RESULTS In the paper, the following parameters will be presented and analyzed as the results of the sensitivity analysis performed by using CERREX code (in comparison with the basic calculation case): • Manpower and cos ts according to the ISDC structure; • Amount of different categories of produced waste. The objective for performing of sensitivity analysis is an impact assessment of input data uncertainties. 4. CONCLUSIONS In the paper, the procedure and methodology for development of decommissioning costing cases for any type of research reactor by using CERREX code are presented by using model calculation example. In the future, there is an intention for updating the calculation procedure by integrating the Inventory database file into the existing structure of the CERREX code; it means to create one robust and compact calculation package used for decommissioning costing of research reactors. REFERENCES [1] INTERNATIONAL ATOMIC ENERGY AGENCY, NUCLEAR ENERGY AGENCY OF THE ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT, EUROPEAN COMMISSION, International Structure for Decommissioning Costing (ISDC) of Nuclear Installations, NEA Rep. No. 7088, OECD, Paris (2012) [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Cost Estimation for Research Reactors Decommissioning. IAEA Nuclear Energy Series, No. NW-T-2.4., IAEA, Vienna (2013) [3] INTERNATIONAL ATOMIC ENERGY AGENCY, Financial Aspects of Decommissioning, IAEA TECDOC 1476, IAEA, Vienna (2005.) [4] PARK. S. K., et al., KOREA RESEARCH REACTOR-2 DECOMMISSIONING PROJECT, ICEM’05:The 9th International Conference on Environmental Remediation and Radioactive Waste Management, Glasgow, Scotland (2005)

Speaker: Mrs Kristina Kristofova (KIND Consultancy, Zvoncin 299, 91901 Sucha nad Parnou, Slovakia)

09:00 BIOOXIDATION PROCESS FOR THE TREATMENT OF LIQUID RADIOACIVE WASTE ARISING FROM THE DECOMMISSIONING OF PHORPHORIC ACID PURIFICATION PLANT OF THE GRESIK PETROCHEMICAL INDUSTRIES

The phosphoric acid purification plant of the Gresik Petrochemical Industries (PAPP-GPI) was operated since February 1989 for uranium recovery from phosphoric acid by two cycle of extraction process using mixed solvent of di-2-ethylhexyl phosphoric acid (D2EHPA) (C16H35O4P) and trioctylphosphine oxide (TOPO) (C24H51OP) in kerosene solution on the weight ratio 4:1:16 for each respectively to obtain the result of pure phosphoric acid and uranium oxide U3O8 or yellow cake. The operation of phosphoric acid purification plant was stopped over 18 years, since 12 August 1989, considering that the facility was contaminated by radiation of uranium and its daughters so the ensure the personnel and public safeties it was necessary to decommosioning the plant by dismantling activity. Decommisioning was performed on 14 October 2004 due to the uneconomically facility operation, unmarketable "yellow-cake" product, the personel and environmental safety consideration, and the wishing of land liberation for the coal project. Decommisioning activities covers the drainning of solution and solid powder of remaining process on the equipment, decontamination of site location and equipment wall, dismantling of equipment, decontamination of equipment after dismantling, and decontamination of concrete floor and wall of building. Before dismantling of equipments, remaining solution of the process and organic liquid radioactive waste were removed from the receiving tank to collecting vessel of treatment at PAPP-GPI area, which has size 14x15x3 m3. The solution in sumpit containing the mixture of raining water and leakage of organic solvent from its tank also to be transfered to the collecting vessel of treatment. The spent chemical solution from decontamination operation as solution of alcohol 80 % in kerosene , phosphoric acid 10 %, and sulfuric acid 10 % for decontamination of equipment where transfer also to the cokkecting vessel. The sollution accumulated in the vessel has the about volume 390 m3, pH 3.48, BOD 2200 ppm, and COD 31500 ppm, activity of alpha and beta are 1200 and 2600 Bq/litre respectivelly. The waste is indicated the category of hazard and poison material containing the organing matters and radionuclides of uranium and its daugthers of Pb-210, Po-210, Ra-226, Th-228, Pu-239, etc. The waste was netralized and treated by biooxidation process using mixture of aerob bacteria of pseudomonas sp, bacillus sp, arthrobacter sp, and aeromonas sp which was performed by aeration and given the nutrision of nitrogen and phosphor. The bacteria grow to prey the organic matters and then multiply rapidly to form the biomass. The organic matters be decomposed to form of CO2 and H2O. The biomass conducts the biosorbtion of radionuclides so that the biomass active sludge and water supernatan be separated . There are detoxification and decontamination of solution. By the continuing of biooxidation process during 54 days was found 365 m3 of the clearence filtrate (liquid of clear solution) with alpha activity 0.4 Bq/litre and beta activity 2 Bq/litre, COD 51 ppm (quality standard 100 ppm), BOD 21 ppm (quality standard 50 ppm), and non-active sludge 14.4 m3, and activity 3.6 m3 with alpha activity 38 Bq/litre and beta activity 1855 Bq/litre. The non-active of filtrate and sludge ware released to the effluent canal of PAF-PKG, where the activity sludge finally was transported to the Center for Radioacive Waste Technology at Serpong to be treated. Keywords : decontamination, dismantling, decommisioning, biooxidation treatment.

Speaker: Mr Zainus Salimin (Center For Radioactive Waste Technology)

09:00 Decommissioning and Reuse of Facilities for Radioactive Waste Management at the UJV Rez, a. s.

Abstract: After 60 years of activities of the ÚJV Řež, a. s. in the nuclear field, there were obsolete nuclear facilities to be decommissioned. Many of these facilities were used for radioactive waste (RAW) management and some of them were or will be reused after decommissioning. ÚJV Řež, a. s. (further also ÚJV) is a leading institution in all areas of nuclear R&D in the Czech Republic and has had a dominant position in the nuclear programme since it was established in 1955 as the Nuclear Research Institute Řež. The activities of ÚJV encompass nuclear physics, chemistry, nuclear power and many other topics. The main issues addressed at ÚJV in past decades have included research, development and services for NPPs, the fuel cycle and irradiation services for research and development in the industrial sector and medicine. ÚJV also manages most institutional radioactive wastes (RAW) produced in the Czech Republic (approx. 90 %). 1. INTRODUCTION The decommissioned facilities for RAW management comprise: - Pipeline system for transport of liquid RAW to a processing facility; - Liquid RAW storage tanks; - Technology for RAW processing; - Facilities for storage of solid RAW, incl. stored RAW. - Decay tanks used for storage of solid RAW with higher activity; Some of the facilities were or will be reused after decommissioning. The facilities were put into operation in the sixties of the last century and were equipped with various technologies for RAW management (RAW stores, liquid RAW storage tanks, evaporators, cementation units, bitumination units, etc.). The technologies have changed over the years, but almost no new technology has been installed in the past, and no substantial repairs of the buildings have been performed since they were put into operation. 2. DECOMMISSIONING OF FACILITIES Preparations for decommissioning began in 1996. Safety analysis report was performed comprising the identification and characterization of potential sources of risk, potentially exposed receptors and exposure pathways, potential chemical compounds, radionuclides and media of concern. The results of the safety analysis report enabled to determine the priorities of the decommissioning, preparation of the decommissioning project as well as the estimation of the expenses. The decommissioning was divided into two phases according to the level of risk and the financing limitations. The total amount of RAW resulting from decommissioning for processing will be approx. 1500 m3, approx. 240 Mg of RAW are expected for release into the environment after decontamination. To meet the legislative requirement, special equipment for measurement is being used. The standard system of solid RAW processing consisting of segmentation and conditioning by cementation into 200 or 216 l drums is combined with disposal of segments of contaminated technological equipment in bigger disposal units; it is advantageous from the point of view of radiation protection because it requires less segmentation operations and it is also less time consuming and many resources are being saved. The Ministry of Finance of the Czech Republic is financing the decommissioning. 3. REUSE OF FACILITIES Because of the high cost of constructing new RAW management facilities, it was decided to reuse/reconstruct the existing facilities (financed by ÚJV). Within the reconstruction of the facilities, new RAW management technologies have been, or will be, installed. Figure 1 shows an example of a decommissioned facility; figure 2 shows a new facility after its installation. Figure 1. Old evaporation unit. Figure 2. New evaporation unit. 4. CONCLUSIONS Decommissioning of nuclear facilities in ÚJV is the only ongoing decommissioning project in the Czech Republic. Decommissioning started in 2003 and will be finished in 2016. Almost all facilities have already been successfully decommissioned. The decommissioning activities are being carried out on the high level of radiation protection and up to now there was no extraordinary event or accident. REFERENCES [1] CH2M HILL United Kingdom, CH2M HILL Česká republika, s r.o., KAP spol. s.r.o., Safety Analysis Report – Decommissioning of Old Facilities in UJV Rez (1997). [2] UJV Rez, Project of Decommissioning of Old Facilities in the UJV Rez (2002). [3] UJV Rez, Update on the Project of Decommissioning of Old Facilities in UJV Rez (2005). [4] PODLAHA, J, Decommissioning of Obsolete Nuclear Facilities in the ÚJV ŘEŽ, a. s., Eastern and Central Europe Decommissioning (ECED) Conference, June 18 - 20, 2013, Trnava, Slovakia, (2013).

Speaker: Mr Josef Podlaha (UJV Rez, a. s.)

09:00 LESSONS LEARNED FROM 30 YEARS OF NUCLEAR DECOMMISSIONING AND ENVIRONMENTAL REMEDIATION

Abstract: AECOM is an international leading provider of fully integrated engineering, construction, programme management, operations, nuclear decommissioning, waste management and technical services to public and private clients. AECOM, through its dedicated strategic business unit “Nuclear & Environment” has been managing the operations and clean-up of high hazard, complex nuclear sites for over 30 years - this has included the decommissioning of 20 reactors, cleanup of 1,290 waste sites and the operation of the only licensed deep geological repository in the world. This paper draws on the vast experience we’ve gained over 150 “site-years” of managing complex, high hazard sites. After a brief description of AECOM, the paper highlights some of our major representative projects involving decommissioning of production and commercial reactors as well as nuclear fuel cycle facilities, clean-up and remediation of nuclear and hazardous sites, and waste management. A wide range of challenges, either common or unique to the sites, generated by routine operations, off-normal events or accident situations, has allowed AECOM to accumulate a wealth of lessons learned. A summary of these are in the final part of the paper under the theme Experience Leads to Improved Performance. 1. INTRODUCTION AECOM’s Nuclear and Environment business provides programme management; planning, design and engineering; systems engineering and technical assistance; construction and construction management; operations and maintenance; environmental remediation; waste management and decommissioning, dismantling and closure services to a broad range of clients, including the Nuclear Decommissioning Authority (NDA) in the UK and the US Department of Energy, as well as privately owned utilities. Currently, AECOM is the lead organisation responsible for managing annual expenditure of over €4.8 billion on projects in the UK and US, has saved its customers around €2 billion and has an excellent industrial and nuclear safety record. 2. PROJECTS In this section, the paper outlines a selection of major projects that have contributed to AECOM’s experience including the River Corridor Closure Project (US), East Tennessee Technology Park (US), Sellafield Sites (UK), the Waste Isolation Pilot Plant (US), the UK’s Low Level Waste Repository, and Decommissioning of the “ThreeYankee” reactors (US). AECOM’s general decommissioning capabilities and knowledge is also described. 3. EXPERIENCE LEADS TO IMPROVED PERFORMANCE The last section of the paper discusses lessons AECOM has learned over three decades of managing its sites, which it has embedded in its approach and now brings to bear on its current projects to support safe, reliable and cost effective delivery. These highlight the need for a relentless focus on safety; the importance of understanding the required end state and “starting with the end in mind”; the need to move from the an operations approach to the discipline of managing a project with a decommissioning mindset; the value of experienced subject matter experts in developing the technical approach and optimised decommissioning plan; the benefit a fully underpinned cost and schedule baseline provides; the recognition that in decommissioning most items end up as waste so an optimised waste disposition plan is essential; and how crucial continuous and open engagement with the full range of stakeholders is.

Speaker: Mr Bonner Robert (AECOM)

09:00 LESSONS LEARNT IN METALLIC MATERIALS CLEARANCE PROJECTS

The operational and decommissioning waste minimization can be achieved by appropriate management of residual materials. Two main categories are dominant in non-activated residual materials, one of them is the metallic materials the other is the building debris. The availability of specific clearance standards (i.e. clearance levels) [1] allows to define, plan, execute and closure comprehensive projects to deal with those contaminated materials. The conceptual approach is based on use of a derived quantity namely Residual Activity Index (RAI), systematic use of Data quality Objectives in different characterization stages, selection of appropriate measurement equipment to segregate and sentencing the materials and a set of Decision rules based on non-parametric tests of hypothesis. All this elements has been tested and implemented in different projects permitting to clear more than 1000 tons. Finally the data quality analysis has permitted to validate the non parametric hypothesis testing used in a cost-effective way and to demonstrate that probability distributions are contaminated distributions because the use of limits of detection of the measurement equipment.

Speaker: Mr RAFAEL GARCIA-BERMEJO FERNANDEZ (IBERDROLA INGENIERIA Y CONSTRUCCION SAU)

1 IAEA_FORM__B_RGFE.pdf

3 IAEA_FORM_A_RGFE.pdf

Paper DER2016_paper_RGFE1.docx

09:00 Optimizing the management of the radioactive waste arising from the decommissioning of the IFIN-HH research reactor WWR-S

Decommissioning of a research reactor is always a challenge for any organization in charge. Planning, an appropriate chooses of strategy and technology, equipment acquisition and waste management are key elements in successfully accomplishing of the decommissioning activities. Optimization of the radioactive waste management by appropriate segregation of the radioactive materials, decontamination, reducing the size, and correct packing are a few methods handy available for every organization who is developing decommissioning operations on its own site. This paper will present some work with regard to radioactive waste management optimization performed in IFIN-HH during decommissioning of its WWR-S research reactor.

Speaker: Mr Angelo-Cristian Dragolici (IFIN-HH)

09:00 Potential for Recycling of Metals within Nuclear Sector in Slovakia

POTENTIAL FOR RECYCLING OF METALS WITHIN NUCLEAR SECTOR IN SLOVAKIA HRNCIR, Tomas; ZACHAR, Matej; ONDRA, Frantisek; DANISKA, Vladimir DECOM, a.s., Sibirska 1, Trnava, Slovakia E-mail address: hrncir@decom.sk Abstract: The decommissioning of nuclear installations represents a complex process resulting in the generation of large amounts of waste materials containing various concentrations of radionuclides. Selection of an appropriate strategy of management of the mentioned materials strongly influences the effectiveness of decommissioning process keeping in mind financial, health and environmental impact. In line with international incentives for optimization of radioactive material management, concepts of recycling and reuse of materials are widely discussed and applications of these concepts are analysed [1-4]. Recycling of some portion of these materials within nuclear sector (e.g. scrap metals or concrete rubble) seems to be highly desirable from economical point of view and may lead to save some disposal capacity. However, detailed safety assessment along with cost/benefit calculations and feasibility study should be elaborated in order to prove the safety, practicality and cost effectiveness of possible recycling scenarios. Paper discussed the potential for recycling of metals generated during decommissioning of NPPs along with approach and lessons learned from previous related research done in Slovakia. 1. INTRODUCTION Since, two NPPs are currently under decommissioning in Slovakia (V1 NPP – standard operation, A1 NPP – shut down after an accident), amount of various waste materials arising from decommissioning is increasing. Therefore identification of potential waste streams, modification and the optimization of waste management system is one of the top priorities in the next few years. Few specialized waste treatment facilities are designed and constructed along with plans for building of melting facility. Moreover, implementation of Council directive 2013/59/EURATOM, laying down basic safety standards for protection against the dangers arising from exposure to ionising radiation, into Slovak legislation will result into stricter clearance levels for several radionuclides of concern. This may lead to decrease of amount of materials that may be free released into the environment and may have impact on overall costs of decommissioning, i.e. the less free released material, the higher waste management costs. Taking into account possible impact of above mentioned factors, potential for clearance, recycling and reuse of these materials should be deeply analysed. 2. APPLIED METHOD AND SELECTION OF RECYCLING SCENARIOS Method for the overall assessment including safety and economical aspects was developed within research project called CONRELMAT devoted to evaluation of possible scenarios for reuse and recycling of decommissioning materials both within and outside of nuclear sector [5]. Applied method follows international recommendations related to the basic safety principles and implements the lessons learned from numerous decommissioning costing calculation cases including simulation of material flow. The principal scheme of the overall assessment of particular scenarios analysed in the CONRELMAT project is depicted in the figure below. Figure 1. Principal scheme of research project “CONRELMAT”. Selection of appropriate metal recycling scenarios should follow one of these principles: • metal material would be rather a shielding than a source (e.g. drums or containers for waste containing higher amount of radioactivity); • radioactivity of recycled metals is very low comparing to the other sources in the vicinity (e.g. steal lining of rooms, tanks for liquid radioactive waste); • metal material would be shielded, used as a reinforcement, i.e. it would be long-term fixed inside of robust structure (e.g. reinforcement bars, or mesh used in construction of larger structures – disposal vaults). 3. CONCLUSIONS Based on calculations performed within the CONRELMAT project, some of the reuse and recycling scenarios are feasible and may save financial resources and disposal capacity. To conclude, paper described preliminary proposal of several possible recycling scenarios within Slovak nuclear industry. However, detailed parameters of particular feasible recycling scenario are inevitable in order to perform more specific safety assessment as well as cost/benefit analysis. Moreover wider discussion including stakeholder involvement (regulatory body, operator, TSO, etc.) is desirable. REFERENCES [1] INTERNATIONAL ATOMIC ENERGY AGENCY, Managing low radioactivity material from the decommissioning of nuclear facilities, IAEA Technical Report Series No. 462, IAEA, Vienna (2008). [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Application of the concepts of exclusion, exemption and clearance, IAEA Safety Standards Series No. RS-G-1.7, IAEA, Vienna (2004). [3] EUROPEAN COMMISSION, Practical Use of the Concepts of Clearance and Exemption, Radiation Protection 122, EC, Luxembourg (2000). [4] U.S. NUCLEAR REGULATORY COMMISSION, Radiological Assessments for Clearance of Materials from Nuclear Facilities, NUREG – 1640, NRC (2003). [5] HRNCIR, T et al., The impact of radioactive steel recycling on the public and professionals, J. Hazard. Mater. 254/255 (2013), 98-106

Speaker: Dr Tomas Hrncir (DECOM, a.s.)

09:00 THE BENEFIT OF ALIGNED POLICY IN THE MANAGEMENT OF LOW LEVEL WASTE

Dennis G. Thompson Managing Director, Low Level Waste Repository, Ltd. AECOM Nuclear & Environment, United Kingdom dennis.thompson@aecom.com Abstract The nuclear industry, like most industries, finds itself with a host of conflicting commercial and regulatory drivers that often lead to less than optimal outcomes. In the 1980's and 90's and into this century, the management of low level nuclear waste in the United Kingdom found itself driven by strong commercial drivers and a changing regulatory regime. By 2006 the rate of waste being disposed in the only national repository was quickly outpacing the capacity of the site, and a forward look at the decommissioning waste inventory indicated that a new Low Level Waste site would need to be located, constructed, and permitted. At that time, many estimates calculated that the site could be full as early as 2030. This paper will describe how several things came together starting in 2006 which enabled the Low Level Waste programme in the UK to make dramatic steps to solve the crisis. I THE RIGHT FOUNDATION The first section of the paper will describe how the UK government effectively set the stage for a successful programme by developing and issuing a national policy for the management of low level waste. In addition to this, government took action to separate the national repository from Sellafield, and set up as a separate and independent Site License Company (known today as the LLWR Ltd.) under the regulatory oversight of the Environment Agency (EA) and the Office of Nuclear Regulation (ONR). Finally a procurement was run to find a parent company that would manage the newly formed license company with a remit to develop and run a National Programme for the Management of Low Level waste based on the recently issued LLW Policy. II THE RIGHT CONTRACT AND THE RIGHT CONTRACTOR In March 2008, a contract was awarded to a consortium led by AECOM and including Studsvik and Areva. This section of the paper will describe how the contract was set up and demonstrate how it was incentivised to align with the government policy. With this expanded remit, and the alignment between contract and policy, the new parent company quickly took the lead in developing the UK Low Level Waste Strategy which received parliamentary approval in 2010, essentially turning the policy and the strategy into law and placing responsibilities for its execution on key players within the UK nuclear industry. At this point the typical tensions between commercial and regulatory drivers started to converge into a single unified national strategy; there was now alignment between government policy, the government backed national strategy, and the commercial contract to operate, not just the Repository, but to manage the national programme. The stage was set to enable the new AECOM led consortium to fundamentally change the industry. This section will further define how the LLWR expanded its service offering from the single service of the disposal of LLW to a wide array of services ranging from metal melting, to characterisation services, to packaging and transport. Around this, a unique set of programme management arrangements were developed to orchestrate the waste producing sites, the repository, and the supply chain to ensure the government’s strategy was effectively delivered on the ground. III THE RESULTS The results are nothing less than dramatic. Just 8 years after the placement of the commercial contract and 6 years after the issuance of the national strategy the fundamental capacity issues that plagued the low level waste industry in the UK have all but vanished. In 2009, virtually all the low level waste generated (95%) was disposed at the Repository. Today the tables have been turned – the strategy is being delivered. More than 85% of the waste generated is now treated and diverted, with just 15% of the waste disposed of at the Repository resulting in annual savings in excess of £50 million across the UK. However, by far the largest benefit is that the UK is no longer concerned about siting a new Low Level Waste Repository site; latest estimates, using the current rates of waste diversion, indicate that the LLWR will not reach its capacity until well into the next century, saving the UK taxpayer the billions of pounds required to bring a new nuclear disposal site into existence not to mention the political challenge of finding a local community willing to accept it. IV. REFERENCES 1. Nuclear Decommissioning Authority UK Strategy for the Management of Solid Low Level Radioactive Waste from the Nuclear Industry: Strategic Environmental Assessment Non-Technical Summary of Environmental and Sustainability Report – June 2009 2. Nuclear Decommissioning Authority UK Strategy for the Management of Solid Low Level Radioactive Waste from the Nuclear Industry: August 2010

Speaker: Mr Dennis Thompson (AECOM)

09:00 THE IMPACT OF STEAM GENERATOR DISMANTLING SCENARIOS ON THE RADIOACTIVE WASTE DISPOSAL

The dismantling of steam generator used in Slovak nuclear power plant V1in Jaslovske Bohunice is planned during the second decommissioning stage (2015 – 2025). In order to optimize the whole dismantling process together with the management of the resulting radioactive waste, the different scenarios were developed. After calculation of external exposure during transport and final disposal of the radioactive waste (carried out in VISIPLAN 3D ALARA code) it is possible to propose the optimal scenario. The results of the calculations show that the regulatory limits are not exceeded in any case.

Speaker: Mr Martin Hornacek (Slovak University of Technology in Bratislava)

09:00 WASTE MANAGEMENT DURING THE REFURBISHMENT OF THE CIEMAT NUCLEAR HOT-CELL

The PIMIC project (acronym in Spanish of “Internal plan for CIEMAT Facilities Refurbishment”) has the main objective of recovering spaces and rehabilitate areas impacted by past activities related to the research and development of peaceful uses of nuclear energy. One of the nuclear facilities to be dismantled was the “Metallurgical hot cell”, where different prototypes of fuel elements were physical - chemically characterized after being irradiated both in JEN experimental reactor and in some commercial Spanish NPP (LWR). This paper describes the dismantling activities and the waste management of this facility. The release of old nuclear facilities at CIEMAT will permit conventional uses of these spaces with no restrictions. It should be emphasized the importance of the clearance process from the economic and environmental point of view.

Speaker: Mr José Carlos Sáez Vergara (CIEMAT)

09:00 → 11:00 Session 5B - 1: Waste Management and Case Studies in Environemental Remediation - Parallel Session

The purpose of this session is to discuss the integration of environmental remediation and waste management activities from initial planning to final waste disposal.

09:10 In-situ Management of Radioactive Material in the UK

In common with any industrial restoration project, decommissioning nuclear sites involves both removing the redundant plant and buildings and remediating the land. If contamination is present in buildings and plant it is likely to be necessary to treat it as radioactive waste for disposal. Similarly, contaminated material in the ground can be excavated for disposal. However, this can be disruptive and very costly, particularly where there are large amounts of soil, rock or below-ground structures that have very limited amounts of radioactivity. Furthermore, the disposal facility is quite likely to be a landfill, which does not offer substantially greater containment of the radioactivity than simply leaving it in place. An alternative approach is to manage the material in-situ. This might involve placing some additional protection over or around the material, which could be combined with the process of landscaping the restored site. The residual radioactivity could then remain under surveillance for a period of time while the hazard decreases due to radioactive decay, and costly space in national radioactive waste disposal facilities would be freed up. The decision on whether this is appropriate is framed by two questions: * Is it sensible to excavate the material and transport it for disposal elsewhere if the risk it poses is very low? * What concentrations of radioactivity could remain in-situ and meet the relevant safety targets? The first question is one of optimisation. In general terms, as NDA conclude, the answer to the question would appear to be “no” for material with very low concentrations of radioactivity. But a site-specific study will always be needed. The second question is one of proportionality. The problem is that situations that are very unlikely, but cannot be completely discounted, drive down generic concentration limits to a point at which in-situ management appears to have limited applicability. However, regulatory guidance for radioactive waste disposal provides alternative criteria for such scenarios. If these are applied then in-situ management has a much wider scope for application. The importance of these points is emphasised by actual reported estimates of radionuclide concentrations in contaminated soil. A simple analysis of four major waste streams in the 2013 UK Radioactive Waste Inventory indicates that all would fail to meet simple and cautious screening values for in-situ management. The costs for disposing of these wastes alone could run into billions of Pounds. The same analysis shows that if more realistic criteria are used then most, possibly all, could be managed in-situ with substantial cost savings.

Speaker: Mr James Penfold (Quintessa Limited)

09:35 OPIMIZATION OF EXPOSURE AND WASTE MANAGEMENT FOR DIFFERENT ACCIDENTAL RELEASES

This paper presents an overview of accidental releases of fission, fusion reactors and cycle installations, (i.e. Iodine alone, Iodine and cesium, iodine plus cesium and short lived airborne nuclides for fission, tritium for fusion and plutonium 239 for fuel installation), and their consequences on the surface deposits management. It addresses the question of dose levels, the effect of some fundamental environmental parameters, the information policy, and the need for a “as rational as reasonably achievable” approach based on efficient and non-expensive actions of the population. Ref. R COULON, J. DELMAS, Ph. GUETAT, C. MADELMONT, R. MAXIMILIEN - Agriculture, environnement et nucléaire, comment réagir en cas d'accident, FNSEA, CNIEL 1991. Ph. GUETAT, P. ARMAND, A. CHARLIER de CHILY, A. FLURY-HERARD, F. MENESTRIER, P. FRITSCH, L. BION - Iodes radioactifs et impacts environnemental et sanitaire : étude bibliographique et quantification – Rapport CEA –R-6065 2004 Philippe GUETAT, Marguerite MONFORT, Eric ANSOBORLO, Valérie MOULIN, Pascal REILLER, Annabelle COMTE, Anne FLURY HERARD, Paul FRITSCH, Florence MENETRIER, Impacts environnemental et sanitaire des isotopes du plutonium : étude bibliographique et quantification - Rapport CEA-R-6186 - Mai 2008 Ph. GUETAT et Al - Plutonium in the environment: key factors related to impact assessment in case of an accidental atmospheric release Pu Future 2008 Radiochemical Acta 97, (2009) 257-260 IAEA – TECDOC 1738 Transfer of Tritium in the Environment after Accidental Releases from Nuclear Facilities (2014)

Speaker: Mr Philippe GUETAT (CEA)

10:00 DIVERSE CHALLENGES ASSOCIATED WITH LEGACY NEAR-SURFACE WASTE DISPOSAL SITES

Numerous countries around the world are currently managing various types of sites contaminated with radioactive materials derived from past nuclear activities and accidents. Many of these sites consist of a shallow excavation in the ground (with or without engineered containment) into which radioactive waste materials were placed - these are sometimes referred to as “legacy trench sites”. Many of these sites were operational in the decades following the Second World War, when research into nuclear power and related activities was rapidly expanding in many countries worldwide. During this period, there was no international consensus on disposal of radioactive waste, and shallow burial was a commonly used method for disposing of low-level wastes. It is now recognized that some of these sites may pose immediate or future unacceptable radiological risks to members of the public and/or the environment, and therefore require the consideration of management options and evaluation of remedial actions. Problems associated with managing these sites include the lack of site-specific information (including characteristics of buried wastes), unclear responsibility and ownership, limited availability of technical expertise, scientific uncertainty, societal issues, and various constraints and limitations. The future management of these sites would be improved by international cooperation aimed at encouraging IAEA Member States to take relevant measures in management of legacy sites. The Legacy Trench (“LeTrench”) initiative, within the IAEA Environet network, aims to address the needs identified above. The activities of this working group will include: developing a global inventory of legacy sites, encouraging preservation of relevant information, identifying and addressing common issues, developing methods of disseminating knowledge, provision of technical assistance or advice, and maintaining relevant expertise.

Speaker: Dr Timothy Payne (Australian Nuclear Science and Technology Organisation)

10:25 MULTIPLE SMALL SIZE SUBSURFACE SOIL CONTAMINATIONS ON AN OPERATING SITE – TOO EARLY TO PREPARE THE FUTURE?

In 2014, during the preparation of a dismantling, contamination in subsurface soil around a waste liquid effluent channel was discovered. This was a stark reminder of the necessity to define a clear policy for all on–site interventions. Presently, the SCK•CEN site is operational and its final end state is not defined. Nevertheless, the protection of the environment (with special attention to the relatively shallow groundwater) and workers carrying out fieldworks in potentially slightly contaminated area’s needs to be guaranteed by using a risk based approach. Furthermore, we believe in the benefit of gathering all data’s to have a solid basis for the problem definition of the remediation when the final end state will be defined. A consolidated strategy has been established covering the following aspects:  Registration of all radiological measurements performed on soils, in a centralized mapping tool in order to design a Conceptual Site Model.  Implementation of dose impact studies for the present soil contamination and the potential ground water contamination.  Enforcement of a strict procedure for the execution of fieldworks on the SCK•CEN site (e.g. Health Physic needs to confirm the absence of contamination before the start of any fieldwork).  Guarantee the workers’ safety during the execution of works in potentially contaminated soil. Some works in such areas cannot be avoided. Recently, two different operations took place in confirmed contaminated soil and a third intervention is foreseen in the beginning of 2016.  Search for new/unexploited removal routes for contaminated soil.  Evaluation of feasible decontamination methods. The experience clearly shows that, at present, we need to:  Avoid excavation when there is a very limited risk/impact for human and the environment. Excavation might not be optimal towards the ALARA principle since no effective removal route is currently available.  Organize protective measures for workers and the environment.

Speaker: Ms Isabelle Majkowski (SCK-CEN)

09:00 → 13:30 Session 5B - Poster

09:00 CALDAS URANIUM MINING AND UOC PRODUCTION SITE: CHALLENGES ON ENVIRONMENTAL REMEDIATION

The Unidade de Tratamento de Minérios (UTM, Ore Treatment Facility) was the first uranium ore mining and ore production site to operate in Brazil. Twenty years after ceasing operations, the site is going through an environmental restoration and decommissioning process, which is to be conducted by the operator and regulated by the national nuclear and environmental authorities. The UTM case can be identified as a typical uranium mining operation carried out before restrictive licensing procedures and environmental laws became a trend around the world. This paper aims to present the current status of the site and some of the challenges to be faced by the parties involved in the decommissioning process.

Speaker: Mr Raul Villegas (Brazilian Nuclear Energy Comission)

09:00 Environmental remediation experience for the long term storage facility of radioactive waste in Cuba.

ABSTRACT: From the need for Cuba to have a facility for long term storage of radioactive waste, the study of the conditions of the existing waste storage for short and medium life was performed. A methodology that included site characterization (including geological, hydrogeological, socio-economics), safety assessment and evaluation of the state of construction of the facility was used for this purpose. Key results showed the existence of problems such as the presence of cracks in concrete joints, which are associated water filtration, high porosity, moisture and corrosion, among others. The need to implement remedial and corrective actions for adapting the facility for the intended purpose was evident.

Speaker: Mrs Gema Fleitas (Center of protection radiationa and hygiene)

09:00 NATURAL RADIATION MONITORING IN THE GOLD MINING, URANIUM, AND THORIUM REGIONS OF CAMEROON: FROM MEASUREMENTS, DOSE ASSESSMENT TO RADIATION PROTECTION

The present paper summarizes the findings of studies carried out in the gold mining, uranium and thorium bearing regions of Cameroon. It also underlines future prospects to strengthen the radiological protection of members of the public exposed to environmental natural radiation in Cameroon. After soil and foodstuff sampling, α- and γ- spectrometry were used to determine activity concentrations of natural radionuclides in these samples. Moreover Electret Ionization Chambers (EPERM) were deployed in the uranium region of Poli and the thorium region of Lolodorf to measure radon in houses. Passive integrated radon–thoron discriminative detectors (RADUET) were used in the high natural radiation areas of Poli and Lolodorf to measure simultaneously indoor radon (222Rn) and thoron (220Rn). 20% of houses in Poli and 50% in Lolodorf have radon concentrations above the reference level of 300 Bq.m-3. 30% of dwellings have thoron concentrations above 300 Bq.m-3 in the high natural radiation areas of the thorium region of Lolodorf. The contribution of indoor thoron to the total inhalation dose in the thorium region of Lolodorf ranges between 15%- 78.5% with the mean value of 47% showing that thoron cannot be neglected when assessing radiation dose. Radioactivity measurements in soil samples collected in the gold mining region of Eastern Cameroon do not prove any environmental pollution by radioactive materials. Thus soil collected from the mining sites can be used as building materials. A comparison of annual excess risk for radon-induced lung cancer with national lung cancer incidence data allows concluding that the local risks are elevated but not necessarily representative of the country as a whole. The above results highlight the importance to put in place strong regulation in Cameroon in terms of radiation protection to better manage exposure of the public to natural radiation stemming from NORM, radon and thoron. Only if this action is made, members of the public will be protected in case of future uranium and thorium mining activities in Cameroon.

Speaker: Dr - SAÏDOU (Nuclear Technology Section, Institute of Geological and Mining Research)

09:00 RADIATION RISK ASSESSMENT OF AN OLD URANIUM MINE SITE, FORTE VELHO, IN PORTUGAL

Abstract: The former uranium mine site of Forte Velho, Guarda, was abandoned in the 70s. Although kept under surveillance till present, there has been no action about the cleanup of this site. A radiation risk assessment of Forte Velho site was carried out on 2014 and results are summarized herein. Radiation dose levels and radioactive waste piles recommend clean-up and implementation of remedial measures at this old uranium mine site.

Speaker: Prof. Fernando P. Carvalho (Instituto Superior Técnico/Laboratório de Protecção e Segurança Radiológica,)

09:00 REMEDIATION OF WISMUT´S URANIUM TAILINGS PONDS – APPROACH AND LESSONS LEARNED OVER TWO DECADES

In Eastern Germany uranium mining lasted from 1945 until 1990 leaving enormous environmental impacts. From 1951 to 1990 the former Soviet-German WISMUT company produced a total of about 216,000 t of uranium, mainly in two large mills located near Seelingstädt (Thuringia) and Crossen (Saxony). Nearby the former Seelingstädt mill tailings dams were erected in mined-out open pits for separate disposal of uranium mill tailings from soda alkaline and from acid leaching processing schemes. Nearby the Crossen mill only tailings from soda-alkaline leaching were discharged into tailings ponds erected in ancient gravel pits and valleys. The tailings ponds Culmitzsch A and B, Trünzig A and B near Seelingstädt and the tailings ponds Helmsdorf and Dänkritz 1 near Crossen cover a total area of about 580 ha and contain more than 160 million m3 of uranium mill tailings. Initially after reunification of Germany the federal government funded in 1991 in total 6.4 billion EUR to cover for decommissioning and remediation of the uranium mining sites in Thuringia and Saxony. Based on the federal Wismut act the state-owned Wismut GmbH was established and made responsible for decommissioning and remediation of the uranium mining sites. First decommissioning activities started in 1991 including defence measures against immediate hazards, environmental investigations and development of initial site-specific remediation concepts. Decommissioning of the uranium tailings ponds has to meet the following most relevant decommissioning aims: • the radioactive dose to the population caused by the uranium mining sites and by decommissioning works is to be limited to no more than 1 mSv/year. • water released from the sites into the hydrographic net must comply with regulatory discharge standards for contaminants; • decommissioning objectives have to be sustained in the long term. This means that sufficient stability and functionality of all relevant constructions, like i.e. soil covers or reshaped dams, shall be ensured for periods of 200 years up to 1,000 years with minimum maintenance and care required in the long term. Wismut identified dry decommissioning in situ as the preferred remediation option for all of its uranium mill tailings ponds in Thuringia and Saxony. Dry remediation option is costly but the risks are significantly lower compared to wet remediation options. By 1996 this was agreed by the authorities of Saxony and Thuringia for all tailings ponds under responsibility of Wismut. Dry remediation in situ includes the following basic decommissioning steps: • removal of pond water, catchment of seepage and contaminated runoff including water treatment before release into the receiving streams; • interim covering of air-exposed tailings surfaces including partial dewatering of soft fine tailings using shallow vertical wick drains; • re contouring of dams with respect to geotechnical stability and recontouring of ponds including the use of deep vertical wick drains and surcharge loading with respect to speed up fine tailings consolidation during the remediation period • final covering including different soil cover designs with respect to the results of modelling of contaminant transport • vegetation of the all surfaces affected by remediation works including balancing of intervention measures and compensatory measures in the accompanying landscape management plan • continuous monitoring during remediation phase regarding radiological, geotechnical, geochemical/environmental aspects and including geotechnical and ecological on-site monitoring of the implementation of the construction works At first the paper presents the approach for the development of the remediation design and the legal, technical and societal constraints. The paper outlines the main aspects of the fundamental remediation steps mentioned above. During more than two decades changing legal, technical and societal requirements affected the design and implementation of each remediation step. The paper presents in more detail legal and technical challenges that led to changes in the technical design of specific remediation steps over the last two decades. In the early remedial stage water management of completely filled ponds were of critical importance. Then the design of interim covering of very soft tailings and dam reshaping with respect to seismic stability had to be solved. During ongoing remediation the recontouring design of pond surfaces changed site-specifically with respect to tailings consolidation and with respect to the experiences made. Different final cover designs have been built or applied for including single layer store-and-release covers, two-layer covers consisting of storage layer and underlying sealing layer and multi-layer type covers. During the last two decades the increasing requirements from environmental law, in particular the law of protection of species and from water law significantly influenced the design, the permitting procedures and the implementation of the works. The paper presents in more detail selected examples that show the particular influence of these changing requirements on the remediation. Finally the paper presents an outlook until the completion of tailings remediation foreseen in 2028 and to the after-care phase.

Speaker: Mr Ulf Barnekow (Wismut GmbH)

09:00 ROMANIAN EXPERIENCE IN REMEDIATION OF NORM CONTAMINATED SITES - case study -

1. Regulatory Framework: • Law 111 on the safe deployment of nuclear activities, republished, stating that for the carrying out of a nuclear activity generating or having generated radioactive waste, the authorisation holder shall compulsorily be responsible for the management of radioactive waste generated by his own activity; bear the expenses related to the collection, handling, transport, treatment, conditioning and temporary storage or final disposal of this waste; pay the legal contribution to the financial sources for the management and final disposal of radioactive waste and spent nuclear fuel and for the decommissioning of nuclear installations. Therefore, the holder of the authorisation issued by CNCAN shall develop a programme for the preparation of the decommissioning and submit it for approval to the Commission and produce the proof of having paid the legal contribution to the financial sources for the management and final disposal of radioactive waste and spent nuclear fuel and for the decommissioning of nuclear installations. - art. 26, 27 • Fundamental Norms of Radiological Safety are seting up the requirements concerning the assurance of radiological safety of occupational exposed workers, population and environment, in accordance with the provisions of Law 111/1996 on the safe deployment of nuclear activities, republished. Chapter VII sets the provisions for regulating the significant increase of exposure due to natural radiation sources, detailing the radiation protection against exposure from terrestrial natural radiation sources. Nevertheless, CNCAN is empowered to carrying out of interventions in order to ensure the implementation of measures with a view to reduce the exposure of workers and/or public, in accordance with the provisions of these norms referring to the interventions. • NORM regulation is drafted 2. Identification of radioactive contamination Whenever the notification is made, CNCAN investigates the area by in situ gamma dose rate measurements and environmental sampling for laboratory analyses. The case study presents the actions taken following an official letter from the Mayor of a Romanian village, claiming a radioactive contamination of the environment, due to some past liquid discharges from gas exploitation in the area. 3. Regulatory dispositions 3.1. Immediate actions required that as soon as the polluter company wass identified, based on the provisions of the Fundamental Norms on Radiological Safety, it was asked: to mark the perimeter of the contaminated area and to limit the access of the public inside the perimeter, to assess the level of radioactive pollution of the contaminated area and to send the results to CNCAN, in order to establish the appropriate remedial actions, if necessary. 3.2. First assessment of radioactive contamination The radiological investigation consisted in: • radiological mapping, performed by systematic measurements of the count rate 10 cm above the ground with different steps along the 3 mapping elements, • measurement of the ambient equivalent dose rate, 1 m above ground, and • field sampling and laboratory analysis by high resolution gamma spectrometry. 3.3. Proposals for remedial actions • removing of vegetation and drainage of water from contaminated area; • removing of a soil layer up to 35 cm thick in areas in which the dose rate exceeds twice the natural background, under radiological supervision; • final disposal of removed material by one of the following methods: in the National LILW Repository or in a sludge pile of the National Uranium Company; slurry of the soil and injection of the mud in an oil extraction well or, the dilution of the contaminant concentration by mixing the contaminated soil with uncontaminated soil on the same location. 3.4. Second assessment of radioactive contamination • new measurements and analyses on the same area, after the removal of vegetation and water , the same methods being used, with slightly different approaches. 3.5. Other remedial proposals • removal of the 35 cm top layer of soil from the hot spots, under radiological supervision; • intermediate storage of the removed material, in one of the Polluter’s facilities, located near the contaminated site; • disposal of the contaminated soil in one of the future disposal facilities designed for hydrocarbon contaminated soil.

Speakers: Mr ALEXANDRU EREMIA (National Commission for Nuclear Activities Control (CNCAN)/ROMANIA), Mrs CORA GOICEA (NATIONAL COMMISSION FOR NUCLEAR ACTIVITIES CONTROL - CNCAN)

09:00 THE REGULATORY CONTROL OF THE DECOMMISSIONING OF PHOSPHORIC ACID PURIFICATION FACILITY, PT PETROKIMIA GRESIK

Abstract Because of economical reason, the operation of Phosphoric Acid Purification Facility-PT Petrokimia Gresik was stopped on 12 August 1989. In June 2000 BAPETEN (Indonesian Nuclear Energy Regulatory Agency) recommended that PT. Petrokimia Gresik apply for decommissioning permit to BAPETEN as soon as possible. BAPETEN had issued the dismantling permit on 2 April 2008. Some criterias had to be met by the company during the decommissioning. During the decommissioning, BAPETEN had conducted 7 inspections at the facility started on 7 April 2008 and ended on 15 August 2008. On 14 April 2009 BAPETEN officially declared that all the inspection findings were closed and on 25 July 2009 BAPETEN issued a Statement Approval of the Site Clearance from BAPETEN regulatory control. 1. INTRODUCTION Phosphoric Acid Purification Facility-PT Petrokimia Gresik was operated in April-July 1989. Because of economical reason, the facility operation was then stopped on 12 August 1989. In June 2000 BAPETEN conducted an inspection at the facility and recommended that PT. Petrokimia Gresik apply for decommissioning permit to BAPETEN as soon as possible. Decommissioning activities covered the draining of solution or solid powder of remaining process on the equipment, decontamination of site location and equipment wall, dismantling of equipment, decontamination of equipment after dismantling, decontamination of concrete floor and wall, and radioactive waste management. Some criterias had to be met by the facility during the decommissioning. The clearance level was 1 Bq/g, 1000 Bq/l for the liquid , surface contamination limit was 1 Bq/cm2, the dose rate at 50 cm from the surface was 0,5 uSv/hour based on BAPETEN Regulation Letter No. 1459A/P 10l/PIBN/2008. Limit values for toxic matter are pH 6-9, COD 100 ppm and BOD 50 ppm (Gov.Regulation of East Jawa No. 45 year of 2002). 2.METHODS PT. Petrokimia Gresik had been given a permit for decommissioning its facility by BAPETEN. The company submitted the report to BAPETEN to comply with all the requirements to ensure that no radioactivity exceeding the clearance level and no contamination exceeding the limits set by BAPETEN. The results of decommissioning process were evaluated by BAPETEN to make a decision that the decommissioning had been performed well to protect the public and the environment. 3.RESULTS After receiving the application from PT. Petrokimia Gresik for dismantling its facility, BAPETEN had issued the dismantling permit on 2 April 2008. The decommissioning resulted in contaminated materials and equipments The radioactive waste was sent to Radioactive Waste Technology Center-BATAN. The non contaminated material and equipment can be reused for reprocessing. Some of non contaminated sludge and all of non contaminated filtrate water which comply to its limit values were released to the effluent release drain system of the plant. During the decommissioning, BAPETEN had conducted 7 inspections at the facility started on 7 April 2008 and ended on 15 August 2008. In August 2008, BAPETEN performed inspections and observed some findings. The facility performed several treatments and corrections to comply with the requirements and limits set by BAPETEN. On 14 April 2009 BAPETEN officially declared that all the inspection findings were closed and on 25 July 2009 BAPETEN issued a Statement Approval of the Site Clearance from BAPETEN regulatory control. 4.CONCLUSIONS The results of decommissioning process were evaluated by BAPETEN to make a decision that the decommissioning had been performed well to protect the public and the environment. The results showed that the radioactivity levels and contamination levels did not exceed the clearance levels and limits. BAPETEN declared that all findings were closed and issued a Statement Approval of the Site Clearance from BAPETEN regulatory control REFERENCES [1]. SALIMIN, Z., ZAID, A., Dekomisioning Fasilitas Pemurnian Asam Fosfat Petrokimia Gresik, PTLR-BATAN (2009) [2]. MIRAWATY, GUNANDJAR, Imobilisasi Limbah Sludge Radioaktif dari Dekomisioning Fasilitas Pemurnian Asam Fosfat dengan Matrix Campuran Bitumen dan Pasir, PTLR-BATAN (2011) [3]. DARYOKO, M., Strategi Dekomisioning Fasiltas Pemurnian Asam Fosfat Petokimia Gresik, PTLR-BATAN (2009) [4]. BAPETEN, Laporan Keselamatan Nuklir 2009 (2010) [5]. BAPETEN, Inspeksi Pelaksanaan Dekomisioning Pabrik Pemurnian Asam Fosfat (PAF) PT. Petrokimia Gresik, (2010)

Speaker: Mr Azhar AZHAR (BAPETEN)

11:00 → 11:30 Coffee Break

11:00 → 11:30 Coffee break

11:30 → 13:30 Session 5A - 2: Optimizing Waste and Materials Management in Decommissioning - Parallel Session

The purpose of this session is:
• To discuss the developments and challenges being faced dealing with materials resulting from decommissioning activities
• To consider the integration of decommissioning and waste management activities from initial planning to final waste disposal (including management of problematic materials)

11:30 → 13:30 Session 5B - 2: Waste Management and Case Studies and in Environmental Remediation - Parallel Session

The purpose of this session is to discuss the integration of environmental remediation and waste management activities from initial planning to final waste disposal.

11:30 Approach for Site Characterization of Historic Waste Management Areas at Chalk River Laboratories

Approach for Site Characterization of Historic Waste Management Areas at Chalk River Laboratories Hogue, Katie Environmental Remediation Project, Decommissioning and Waste Management, Canadian Nuclear Laboratories, Chalk River, Ontario, Canada Email address: katie.hogue@cnl.ca Abstract: Canadian Nuclear Laboratories (CNL), formerly Atomic Energy of Canada Limited (AECL), has been investigating and verifying the waste inventory in the Chalk River Laboratories (CRL) Waste Management Areas (WMAs). CNL carries out characterization and assessment of the WMAs to identify health, safety, and environmental liabilities and ensure compliance with regulations. Characterization results are used to model risks and future site conditions to determine where remediation activities, both in the interim and for final release of the site, are necessary. The characterization results also serve as input to developing the overall management strategy or decommissioning approach for the WMAs. The waste inventory of the historic WMAs is known primarily from historical records, which may be inaccurate or incomplete. CNL has been working to recover as much information as practical on CRL legacy waste and record it in an electronic database. Verification of the waste records is made through field investigations. The characterization work also supports ongoing revision to the site wide Environmental Risk Assessment (ERA) by providing information on the source terms and soil conditions within the non-operating WMAs. CNL has adopted a phased approach to characterization, using progressively more invasive techniques in each subsequent phase. Characterization is broken down into three steps (1) Historical Site Assessment, (2) Non-Intrusive Characterization and (3) Intrusive Characterization. Methods The first step in characterizing any contaminated area is to conduct a Historical Site Assessment (HSA). The purpose of an HSA is to compile all historical information regarding the site’s use, the physical setting, any previous characterization work and develop a preliminary conceptual site model for the area. Non-intrusive field characterization work is directed at defining the physical and environmental conditions at the site without compromising the physical boundaries of waste structures (i.e., radiation surveys, geophysical surveying, soil coring and sampling, groundwater sampling and excavations). This fieldwork can include excavation of the targeted waste structure to collect information on the geometry and condition, but the scope would exclude penetration of the waste container(s) and sampling of the source material. Intrusive characterization involves entry (intrusion) into the physical boundaries of the waste structure. This fieldwork is aimed at confirming quantitatively the characteristics of the wastes, assessing the extent of soil contamination, identifying more precisely the potential hazards to workers and the environment, and enabling the subsequent development of safe procedures for waste recovery, waste conditioning (if needed) and waste dispositioning.   Case Study – Waste Management Area A (Active Liquid Waste Tank) WMA A was the first location used for radioactive waste management in Canada and was in service from 1946 to 1957. It contains unlined sand trenches and approximately 15 discrete waste burial structures. It was also used for direct dispersal of radioactive liquids. Up until the late 1970s WMA A was also used for temporary, surface storage of used equipment. In 1981 the surface was cleared and graded and WMA A was closed to any new additions of waste. Preliminary characterization activities (non-intrusive) were undertaken at WMA A in the 1990s and early 2000s including geophysical surveys and excavations to locate and evaluate the condition of several waste structures of interest. At that time, CNL had identified two waste structures within WMA A that contained liquid. There were also another 15 historic structures in WMA B that contained liquids. Investigations of the structures containing liquids were advanced ahead of other waste burials. The Active Liquid Waste Tanks (ALWTs) in WMA A were two of these structures and the sequence of characterization activities for ALWT 2 is outlined below. In parallel to the advancement of ALWTs characterization and assessment, the overall characterization of WMA A has progressed. Radiological surveys were completed in 2012 and the comprehensive HSA was completed in 2013. In 2014 and 2015, the remaining discrete waste structures within WMA A were investigated. This non-intrusive characterization included excavation, soil sampling and in-situ gamma spectroscopy to examine the construction of waste packages and reduce uncertainty regarding the inventory. The results of the non-intrusive characterization will provide input to ensure the interim safety of WMA A prior to final remediation and will provide a basis for developing the remedial action plan. References [ 1] INTERNATIONAL ATOMIC ENERGY AGENCY, Remediation Process for Areas Affected by Past Activities and Accidents, Safety Series No. WS-G-3.1, IAEA, Vienna (2007). [ 2] INTERNATIONAL ATOMIC ENERGY AGENCY, Characterization of Radioactively Contaminated Sites for Remediation Purposes, IAEA-TECDOC-1017, IAEA, Vienna (1998). [ 3] INTERNATIONAL ATOMIC ENERGY AGENCY, Overcoming Barriers in the Implementation of Environmental Remediation Programmes, Nuclear Energy Series No. NW-T-3.4, IAEA, Vienna (2013).

Speaker: Ms Katie Hogue (Canadian Nuclear Laboratories)

11:55 DECOMMISSIONING OF DAMAGED NUCLEAR FACILITIES AT AL-TWAITHA NUCLEAR CENTER-BAGHDAD -IRAQ.

Qussay, Albustani &Khudhair,Waheeb Iraqi Decommissioning Directorate , Ministry of Science and Technology, AL-Jaderyia street,Baghdad,Iraq. E-mail address: qussayalbostany@yahoo.com : khudhair_57@yahoo.com 1. INTRODUCTION The purpose of participation in this conference is to share knowledge and experience gained in addressing the challenges associated with the decommissioning and remediation of four damaged nuclear facilities at AL-Tuwaitha nuclear center which includes 18 damaged nuclear farcicalities , our case is considered as special case because these facilities have been destroyed by bombing. For safely removal of these facilities , it is require to maintain the communication with experts from IAEA and various member states those who have cases of damaged nuclear facilities and then disseminate and benefit from practical information and lessons learned deriving from these cases. 2. METHDOS Figure (1) shows the decommissioning process flowchart that designated for removal the former nuclear facilities at AL-Tuwaitha nuclear center. 3. RESULTS Table (1) below shows the Master decommissioning schedule,the removal of all formerdamaged nuclear facilities at AL-Tuwaitha center divided into three main phases , scienc wehave no experience in decommissioning,so at the beginning we concentrate on building capacities of our operation staff and then moved them to the decommissioning area starting with low risk facilities those included in phase (1). Phase( 1):Decommissioning of (Geopilot, LAMA, Radioactive Isotope production Facility). Phase(2):Decommissioning planning (Tammuz-2, Adaya, IRT-5000, Fuel Fabrication Facility). Phase(3):Decommissioning of the remaining facilities based on prioritization scheme . 4.CONCLUSIONS For phase(1), all project were accomplished 100% with deviation of one year, . In phase (2) ,Decommissioning of Tammuz-2 reactor and fuel fabrication factory about 65% of the two projects were accomplished . But Adaya was deferred because of security situation and also IRT5000 reactor was differed because of lack in funding. REFERENCES [1] SRS-45.Standard Format and content for safety related decommissioning document, Vienna, July 2005. [2] IAEA safety standards series No. WS-G-5.2: Safety assessment for the decommissioning of facilities using radioactive material,Vienna,2008. [3] IAEA, Characterization of Radioactively Contaminated Sites for Remediation Purposes, IAEA-TECDOC-1017, IAEA Vienna, 1998. [4] IAEA Safety Standards Series Ds298 Draft 5: Fundamentals of Nuclear, Radiation, Radioactive Waste and Transport Safety, IAEA Safety Standards Series Ds298 Draft 5, 2003. [5] Jarjies, A., Abass, M., Fernandes, H., and Coates, R. Development and Application of a Prioritization Methodology for Decommissioning the Iraq Former Nuclear Complex, 2008. [6] IAEA Technical Reports Series No. 463 (2008): Decommissioning of Research Reactors and OtherSmall Facilities by Making Optimal Use of Available Resources, IAEA, and Vienna, Austria. [7] Mahmoud, Tariq A., (1988), Environmental Technology and Science, Ministry of Higher Education and Scientific Researches, Al-Musil University, Iraq, pp. 327-340. [8] IAEA, Clearance levels for radionuclides in solid materials, Application of exemption principles, Interim report for comment, IAEA-TECDOC-8S5, 1996. [9] Safety series No.115 ,international Basic Safety Standards for protection against Ionization Radiation and for safety radiation sources. IAEA ,Vienna,1996

Speakers: Khudhair AL-Jumaili (Iraqi Decommissioning Directorate -Ministry of Science and Technology), Mr QUSAY ALBOSTANI (Ministry of Science and Technology- Iraqi Decommissioning Directorate)

12:20 Chernobyl NPP cooling pond decommissioning

Based on the design, the cooling pond (CP) is a source of technical water intake for the Chernobyl NPP needs. The CP was constructed by creating an artificial embanking dam at floodplain area of Prypiat River with the area of about 22.9 km2. The CP water volume is designed for cooling of four ChNPP Units in the electrical power generation mode. Losses from filtration and losses for the CP water evaporation are compensated from Prypiat River by delivery pumps up to the design level of 111.00 meters on the Baltic system (mBS) that during most part of a year is higher than the level in Prypiat River for 6-7 meters. The CP specificity is its location within the area of Exclusion Zone and Zone of Unconditional (Obligatory) Resettlement - the area radioactively contaminated as a result of 1986 accident at the ChNPP Unit 4. According to the Law of Ukraine “On the Legal Regime on the Territory that Suffered from Radioactive Contamination as a result of the Chernobyl Catastrophe”, the CP area and adjacent territories are covered by the same rules and restrictions that are applied for the Exclusion Zone and Zone of Unconditional (Obligatory) Resettlement. This means that any activity that does not ensure the radiation safety regime is forbidden. R&D and monitoring works targeted at the CP safety assurance, assessment of its radiation and environmental conditions and justification of measures on its decommissioning and further remediation activities can be carried out within the CO water area and adjacent territories with the permission from the ChNPP management and agreement with the Exclusion Zone’s sanitary and epidemiological service. As a result of the beyond-design-basis accident at the ChNPP in 1986, radioactive aerosols and dispersed fuel particles from the destroyed reactor were deposited onto the CP surface. In addition, about 5,000 m3 of contaminated waters from the Plant’s technical supply systems were discharged into the CP. Part of water that was used for fire-fighting at Unit 4 during the accident, as well as partially the waters that were collected in the stormwater sewer system after the ChNPP industrial site decontamination, also arrived there. As of today, the amount of water in the ChNPP cooling pond has started to exceed by many times the ChNPP needs at the decommissioning stage. In so doing, the annual costs for maintaining the safe level of the CP dam and operation of the pump station are about 5 (five) mln. hryvnias. That is why a decision to decommission the CP has been taken. At the present time, the inlet and outlet channels are cut off from the CP, by that creating an independent technical water body ensuring the SSE ChNPP needs in cooling water and water for fire-fighting systems. The nominal level is maintained in this water body by operation of the pump stations from wells. Due to the level lowering in the cooling pond, the SSE ChNPP administration fulfilled organizational and technical activities on radiation protection. The radiation monitoring is arranged along the perimeter of the CP bank line and at dried areas of the CP slopes generated as a result of the water level lowering. Control levels of EDR, activity concentration of -;-LLN (long-lived nuclides), Cs137 and Sr90 in the air along the perimeter of the CP bank line and at dried areas of the CP slopes were established. The CP bank line perimeter was referred to the free regime area, while the exposed areas of the CP slopes – to the tightened control area. The Chernobyl NPP CP Decommissioning Feasibility Study (FS) made forecasts for different scenarios of the controlled lowering of the level in the CP, including a scenario of continuous passive water descent without controlling the water levels’ regime along the CP surface area. The forecasted calculations allow making a conclusion that exceedance of the control levels, in particular pertaining to concentration of radionuclides in the air, can have sporadic character in situation with extreme climatic conditions. In fact, tendencies to the increase and exceedance of the control levels of aerosol activity of -;-LLN, Cs137 and Sr90 in the air along the perimeter of the CP bank line have not been recorded currently. The results of the conducted radiation monitoring, taking into account the forecasted estimates of radiation situation changes given in the CP Decommissioning FS, show the absence of negative impact of the cooling pond slopes’ drying process on the radiation situation in a case of possible further lowering of the water level. It should be noted that a wide network of atmospheric precipitation stations was created within the SSE ChNPP industrial site. The stations are located around the CP within and outside the guarded perimeter of the SSE ChNPP industrial site.

Speaker: Mr Olexandr Novikov (Chernobyl NPP)

12:45 Remediation of a site containing a large amount of contaminated soil: the Flix Dam Project

See attached file.

Speaker: Mr Kaifer M.J. (FCC)

13:30 → 15:00 Lunch break

15:00 → 16:30 Session 6 - 1: Project Management, Skills and Supply Chain Considerations

The purpose of this session is to highlight how experience over the past 10 years has impacted the supply chain and identify future trends – to highlight current good practice in terms of:
• development and implementation of programmes and projects; and
• highlighting needs, opportunities and challenges related to knowledge management, education and training to support future project implementation

15:05 Case Study on experience in organizing and implementing the decommissioning of Ignalina Nuclear Power Plant with two RBMK 1500 Reactor Units

See attached file.

Speaker: Mr Darius Janulevicius (None)

15:00 → 18:00 Session 6 - Poster

15:00 eOMEGA – DEVELOPMENT OF ISDC BASED DECOMMISSIONING COSTING PLATFORM

Abstract: The main purpose of the paper is to present the development procedure of the eOMEGA as a platform for decommissioning costing of nuclear facilities. The eOMEGA costing platform, intended to be a part of the universal eOMEGA platform covering also the other activities within the back-end of nuclear power engineering, is considered to be a tool for assessment and transparent presentation of decommissioning costing data in the harmonised structure. The methodology for cost calculation in the eOMEGA is in line with the international recommendations and best practices [1] and is based on the methodology defined in the International Structure for Decommissioning Costing (ISDC) of Nuclear Installations [2]. 1. INTRODUCTION The eOMEGA platform is a logical follow-up of the DECOM long term involvement (more than two decades) in the field of decommissioning planning and costing including the development of unique decommissioning costing codes. The motivation for eOMEGA platform development is to create the decommissioning costing tool taking into account following assumptions: • fully implemented ISDC structure and methodology to facilitate the harmonisation, transparency and benchmarking in the decommissioning costing; • contribution to promoting, extending or future upgrading of the ISDC; • meet the actual international requirements and trends in costing; • applicable for any type of nuclear facility (power plant, research reactor, waste management facilities, etc.) with any systems, structures and radiological situation; • applicable in any phase of the decommissioning planning process starting from very preliminary phase during construction of the facility up to final planning just before the start of decommissioning; • user should have the costing case fully in “own hands”; • available to any stakeholder involved in the decommissioning planning; • flexible, open, transparent, modern and user-friendly environment; • accessibility via internet and supported by secure internet technologies. 2. METHODS The basic idea behind the eOMEGA platform development is the connection of two existing matured solutions: • Decommissioning costing software OMEGA with fully implemented ISDC structure and methodology; • Web-based ADIOS platform with tools and processes to implement any web-based software solution with user-friendly interface. The core of the eOMEGA platform is a decommissioning costing module with ISDC as a basic calculating structure and format for presenting the costing results. The long term experience with the following specific areas of decommissioning costing are applied in the eOMEGA costing module: • development and implementation of inventory data (physical and radiological) for the decommissioning costing purposes; • determination of input data for decommissioning costing (e.g. unit factors); • calculation of waste management data using unique system simulating the material and radioactivity flow in the decommissioning process; • presenting the costing results in the ISDC format. Except of the decommissioning costing, the following modules are intended to be implemented to the eOMEGA decommissioning costing platform: • specific ISDC costing module for transition period between shutdown of the facility and decommissioning itself; • module for transforming the data from the other structures to ISDC including the benchmarking of the calculation results; • ISDC decommissioning cost risk assessment module. To the future, there is an effort to create an universal eOMEGA platform covering not only the decommissioning costing process but also contains a costing modules for back-end of nuclear fuel cycle, site remediation (including relevant ISDC extensions) as well as the module(s) for funding of all of the above mentioned back-end activities. 3. CONCLUSIONS The paper presents a long term experience and lessons learnt in the field of decommissioning costing. These experience are applied and transferred to the ISDC decommissioning costing platform eOMEGA, which provides the flexible and graded approach for reliable cost estimation explicitly in the recommended ISDC format. The used costing methodology is applicable for any type of nuclear facility and the ISDC costing platform could contribute to greater transparency of decommissioning process and building authorities/stakeholders confidence in the decommissioning cost estimates. REFERENCES [1] INTERNATIONAL ATOMIC ENERGY AGENCY, Financial Aspects of Decommissioning, IAEA TECDOC 1476, IAEA, Vienna (2005) [2] INTERNATIONAL ATOMIC ENERGY AGENCY, NUCLEAR ENERGY AGENCY OF THE ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT, EUROPEAN COMMISSION, International Structure for Decommissioning Costing (ISDC) of Nuclear Installations, NEA Rep. No. 7088, OECD, Paris (2012)

Speaker: Dr Matej Zachar (DECOM, a.s.)

15:00 HOW BENCHMARKING CAN SUPPORT THE SELECTION, PLANNING AND DELIVERY OF DECOMMISSIONING PROJECTS

Abstract: Decommissioning projects are characterized by several risks, long schedule and cost estimates that lie in the range of hundreds of billions of dollars. In many countries, decommissioning projects are even more significant than the nuclear new build. Despite this extremely high relevance there is a huge gap in the literature concerning the benchmarking of these projects. The ultimate goal of the research is to understand communalities between successful projects and unsuccessful ones to draft guideline for the project management. The research methodology presented here is based on a structured process to review the cases and to spot “cross-case” patterns. These “patterns” will then be generalized to generate theoretical propositions. 1. INTRODUCTION Decommissioning projects are characterized by several risks, long schedule and cost estimates that lie in the range of hundreds of billions of dollars. In many countries, decommissioning projects are even more significant than the nuclear new build. Despite this extremely high relevance there is a huge gap in the literature concerning the benchmarking of these projects. The budgets for these projects keep increasing and key stakeholders have a limited understanding of the key determinants that engender these phenomena. 2. METHOD The term “benchmarking” refers to the process of comparing projects and it offers significant potential to identify best practices and improve the performance of project selection, planning and delivery. However, benchmarking of nuclear decommissioning projects is a much debated topic, as it refers to several interrelated subjects. Moreover, it is hindered by the uniqueness, difficulty in gaining appropriate data, and a lack of sufficient data for comparability between projects. Therefore, even if project benchmarking is the envisaged approach to tackle this challenge, it has only been partially used. 3. RESULTS The researchers envisage using an inductive cross-case analysis as in [1] to perform the project benchmarking. This research methodology is based on a structured process to review the cases to spot “cross-case” patterns. These “patterns” will then be generalized to generate theoretical propositions. The approach adopted is based on the seminal work of Eisenhardt [2] who derived a process where theoretical generalizations could be generated from reviewing a set of cases of a particular phenomenon: statistical analysis can then be used to reveal relationships between project characteristic (independent variables) and project performance (dependent variables). Nevertheless, there are inherent problems in trying to understand these relationships, due to: 1) the numbers of fully decommissioned project available, too small for statistical purposes; 2) data associated with decommissioning project characteristics are rich and quantitative and needs to be converted into a quantitative form to enable statistical analysis. This process is notoriously difficult [3]; 3) the evaluation of “performance” for projects in general and PPMs in particular can be controversial [4]. Suggested methodologies to tackle these issues are the Fisher Exact Test and the Qualitative Comparative analysis. Therefore, instead of providing a bottom-up analysis, this research focuses on a top-down benchmarking approach and presents the key ideas of a project to research benchmarking utilizing information from the UK’s Nuclear Decommissioning Authority (NDA) and publicly available documents. Being the benchmarking process is as important as the benchmarks themselves [5], this study is based on the critical selection of the relevant case studies. 4. CONCLUSION This initiative is receiving a particular focus because the estimates for the decommissioning programme are large and increasing, with a current average of £ 53.2 billion ($ 80 billion) for the decommissioning of Sellafield (UK) only, and accounting for more than half of the decommissioning costs within the current NDA programme. Also, many other facilities across the UK are undergoing preparations to be decommissioned, and even if there has been considerable progress with this exercise, there is a need to institute a sustainable approach. The authors proposes an innovative approach to find the relationships between dependent and independent variables, through an inductive process, and envisage a deep reflection on the way forward for the application of project benchmarking in nuclear facilities. The key audience of this presentation include policy makers, regulators, companies involved in the decommissioning, owners of facilities to be decommissioned and the key stakeholders involved in the delivery of decommissioning projects. REFERENCES [1] N. J. Brookes and G. Locatelli, “Power plants as megaprojects: Using empirics to shape policy, planning, and construction management,” Util. Policy, vol. 36, pp. 57–66, 2015. [2] K. M. Eisenhardt, “Building Theories from Case Study Research,” Acad. Mangement Rev., vol. 14, no. 4, pp. 532–550, 1989. [3] M. Easterby-Smith, M. Thorpe, and R. Jackson, Management Research. SAGE Publiscation, 2011. [4] L. A. Ika, “Project Success as a topic in project management journals,” Proj. Manag. J., vol. 40, no. 4, pp. 6–9, 2009. [5] N. Garnett and S. Pickrell, “Benchmarking for construction: theory and practice,” Constr. Manag. Econ., vol. 18, no. 1, pp. 55–63, 2000.

Speaker: Ms Diletta Colette Invernizzi (University of Leeds, UK)

15:00 PERSONNEL TRAINING IN UNIVERSITY FOR DECOMMISSIONING OF NUCLEAR FACILITIES: EXPERIENCE AND PROSPECTS

Abstract: Nowadays the decommissioning of nuclear facilities issue is becoming increasingly important in the world and in the Russian Federation as well. For successful and safe decommission the availability of qualified trained personnel is an essential requirement. As far as NPP decommissioning is developing, it is necessary in addition to retraining and advanced training of employees of the nuclear industry organizations train specialists in decommissioning area as far back as their studying in a university. At the Obninsk Institute for Nuclear Power Engineering of the National Research Nuclear University MEPhI specialist training in decommissioning was conducted as early as in the 2000s. Gathered experience allows now to start personnel training in the NPPs decommissioning. In the report existing experience of personnel training is presented as well as possible options of such training for different work skills and training time. Further some considerations regarding personnel training for foreign countries with Soviet build nuclear facilities are presented. 1. INTRODUCTION Nowadays the decommissioning of nuclear facilities issue (especially with regard to nuclear power plants) is becoming increasingly important in the world and in the Russian Federation as well. In the beginning of the 21st century this process was started for five NPP units only in Russia (the first Obninsk nuclear power plant and the first stages of Novovoronezh and Beloyarsk NPPs). Now on the agenda are decommissioning of four units of Bilibino NPP and 2nd stage's two units of Novovoronezh NPP [1]. For 1st generation RBMK-type units (Leningrad and Kursk NPPs) the problems with graphite are successfully resolved, but decommissioning of several units cannot be excluded. 2. METHDOS The decommissioning is a complex process with technical, organizational and social aspects. For successful and safe decommission the availability of qualified trained personnel is an essential requirement. It should be noted that immediate dismantling of equipment is performed by regular personnel just as delayed dismantling is likely to be carried out by new staff with no operation experience. Given that to date NPPs decommissioning was carried out in a rather limited extent, personnel training was mainly carried out by the nuclear industry organizations such as nuclear power plants, training and engineering centers and so on. As far as this process is developing, it is necessary in addition to retraining and advanced training of employees of the nuclear industry organizations train specialists in decommissioning area as far back as their studying in a university. 3. RESULTS At the Obninsk Institute for Nuclear Power Engineering of the National Research Nuclear University MEPhI specialist training in decommissioning was conducted in the 2000s based on Nuclear reactors specialty. The main emphasis was placed on the addition of courses related to regulatory and legislative acts, radiation safety and technical issues. Gathered experience allows now to start personnel training in the NPPs decommissioning based on specialties named "Nuclear power plants: design, operation and engineering" and "Nuclear reactors and materials." 4. CONCLUSIONS In the report existing experience of personnel training is presented as well as possible options of such training for different work skills and training time. Further some considerations regarding personnel training for foreign countries with Soviet build nuclear facilities are presented. REFERENCES [1] ZIMIN, V.K., KORNEEV, I.I., Decommissioning as a management tool for multi-unit NPP service life, Nuclear and Environmental Safety 3 (2012) 78–80.

Speaker: Mr Alexander Nakhabov (National Research Nuclear University MEPhI)

15:00 → 16:35 Young Professional Session - 1

15:00 → 18:00 Young Professional Session - Poster

15:00 A Study on Ultrasonic Visualization Technique for Leakage and Fuel Debris Inspection (1F).

This study investigated a measurement system combining ultrasonic velocity profiler (UVP) method and aperture synthesis method. The system was devised as an inspection technique to identify leakage point and to determine distribution of fuel debris for decommissioning of Fukushima Dai-ichi nuclear power plant. To evaluate the developed measurement system, verification experiments were conducted in the tank where water is leaking and a stone is placed in the flow field as mock fuel debris. As a result, applicability of the developed measurement system was confirmed.

Speaker: Mr Takuya Kawachi (Dept. of Nuclear Engineering, Tokyo Institute of Technology)

15:00 DECOMMISSIONING OF CESNEF L-54M RESEARCH REACTOR: RADIOLOGICAL PRE-CHARACTERISATION OF GRAPHITE AND BIOLOGICAL SHIELD

Abstract: The L 54M thermal research reactor was commissioned by Politecnico di Milano to Atomic International in 1958. It was shut down in 1979 after around 20 years of operations for research purposes. Subsequently it was put under Safe Storage. Several operations have to be managed in order to restore the reactor site to the status of “unrestricted re-use”, the so-called “greenfield” status. This work concerns the preliminary radiological characterization of graphite and concrete, which are the main constituents of the moderator/reflector and the biological shield respectively. The main purpose of this work is the assessment of conventional and radioactive wastes volumes in view of the final decommissioning. In this context, the key radionuclides deriving from neutron activation are: 3H, 14C, 60Co, 152Eu. The first part of the research was based on neutron diffusion theory for the computational evaluation of the activity of these activation products. The second part of the work was focused on Non-Destructive Analyses (NDA) and DA of few graphite and concrete representative samples [1]. 1. INTRODUCTION The first step of decommissioning consists in the Historic Site Assessment (HSA) in order to acquire a complete outline of the L-54M reactor during its operational life [2]. Thanks to this analysis, the main reactor components involved in the activation process are identified, as reported in Table 1 [3]. Consequently, a preliminary radiological characterisation has to be performed not only to acquire the knowledge on the radiological inventory (fingerprint) but also to quantify the specific activity of the individual radionuclides. Table 1 Main reactor components Components DIMENSIONS Material Mass [kg] Supplementary Shield h = 1616 mm ∅ = 558.8 mm Alluminium Reflector V = ∼ 7 m3 Nuclear graphite ∼ 11000 Core R = 200 mm Smin = 1.73 mm Steel 9.071 Control rods Internal cladding ∅int = 21 mm Alluminium External cladding Steel 0.2336 Irradiation channels Iron ∼ 3200 Fuel Vsol = 27.4 l UO2SO4 + H2O 1.3358 Biological shield ∼ 189 m3 Barite concrete (ρ = 3.66-3.82 gcm-3) ∼ 700000 2. METHODS The computational evaluation was performed by means of Matlab codes supported by Monte Carlo simulations [4]. 60Co and 152Eu specific activities were measured by HPGe gamma spectrometry, while 3H and 14C were analysed by liquid scintillation after suitable sample pre-treatment in a furnace. 3. RESULTS The preliminary results on graphite bricks are summarised in Table 2. The average specific activity of key radionuclides in the three main components are reported in Table 3. Table 2 Main radionuclides in graphite bricks (distances are from the external surface of the reflector) Specific activity [Bqg-1] Depth [cm] 14C 3H 152Eu 0 4.1 ± 0.3 51.5 ± 3.4 4.1 ± 0.3 40 37.5 ± 2.6 811 ± 54 34.3 ± 2.3 120 370 ± 25 9170 ± 597 253.9 ± 18.8 Table 3 Summary of activated materials Material Volume [m3] Mass [kg] Nuclide Activity [GBq] Specific average activity [Bqg-1] Graphite (moderator + irradiation facility) 9.5 16100 3H 37.44 2.3∙103 14C 1.58 98.1 152Eu 1.19 73.9 Barite concrete (estimate of activated quantity) 4.1 15700 60Co 0.09 6.0 152Eu 0.58 37.0 Steel (core ) - 20 60Co 1.97 9.9∙104 4. CONCLUSIONS The analyses performed in bricks at different depths confirmed that the whole graphite is completely activated. Thus it has to be considered as a radioactive waste. Contrariwise, the concrete seems to present only a limited activation in the inner part. REFERENCES [1] IAEA, Technical Report Series no.44622, Decommissioning of Research Reactors: Evolution, State of the Art, Open Issues, 2006. [2] NRC, EPA, DoE, DoD, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), 2000. [3] Sergio Terrani, Il reattore e gli impianti nucleari del CeSNEF: Rapporto di Sicurezza, Politecnico di Milano, 1961. [4] Mario Terrani, Il reattore L54-M del CeSNEF. Misure di flussi neutronici e determinazione della potenza, 1960.

Speakers: Mr Eros Mossini (Politecnico di Milano), Mr Luca Codispoti (Politecnico di Milano)

15:00 DECOMMISSIONING OF L-54M REACTOR: PRELIMINARY TOPSOIL CHARACTERISATION OF REACTOR SITE AND SURROUNDINGS

Abstract: Scientific research cannot disregard environmental and public safety. The majority of worldwide nuclear power plants will undergo permanent shutdown over the next 10 years. This is also the case of L 54M, a homogeneous fuel thermal research reactor commissioned by Politecnico di Milano to Atomic International in 1958. It was used for research purposes in the fields of nuclear reactor physics and control, radiochemistry and for materials irradiation. It was shut down in 1979 and subsequently put under Safe Storage. In order to restore the reactor site to the status of “unrestricted re-use”, also known as “greenfield” status, several operations have to be managed. The subject of this work is the radiological characterization of the topsoil of the CeSNEF site, with the aim of assessing that no external radionuclide release occurred during the operational life of the plant. Standardised methods were employed in order to obtain representative and reliable results. Suitable sample pre-treatment procedures were applied. Gamma and beta spectrometric analyses were carried out. 137Cs and 90Sr were considered as representative radionuclides which could have been originated from reactor operations. 1. INTRODUCTION The topsoil radiological characterization of the reactor Impacted zone is a preliminary step to final decommissioning. It aims at evaluating the radioactivity inventory in the area before the plant dismantling. The present work is based on the guidelines of the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) [1] and the Environmental Radiation Survey and Site Execution Manual (EURSSEM) [2]. They provide a complete guidance on radiological, environmental and facility surveys. They are based on the best available practices for demonstrating compliance with dose or risk-based regulations of radioactively (or potentially) contaminated sites. Depending on the type of the matrix and on the future use of the material to be analysed, clearance levels were established by the Italian Competent Authorities [3]. The key radionuclides of concern when dealing with potential nuclear facility external contamination are: 3H, 14C, 40K, 54Mn, 55Fe, 60Co, 63Ni, 90Sr, 137Cs, 152Eu, 226Ra, 235U, 238U, 232Th, 241Am. Among these, 40K and the progenies of 238U and 232Th are naturally occurring radionuclides in environmental samples. 2. METHODS The area surrounding the CeSNEF site was radiologically defined as Impacted zone of Class 3 [1]. For this area a uniform distribution sampling was chosen, together with a spot-like sampling in the not-Impacted zone [1,2]. The number of samples has to guarantee statistical significance with a standard level of confidence (usually 95%) [1,2,4]. The initial number of samples was set to 30, but supplementary samples will be collected if the statistic constrains are not respected. Georeferenced topsoil samples were collected following a standardised procedure [5]. Before analysis, proper sample preparation was necessary in order to obtain a homogeneous matrix. 250 mL plastic cylindrical beakers were filled with dried (110°C for 24 h) and sieved (up to 2 mm) samples for the HPGe (ORTEC GEM) gamma spectrometry measurement [5,6]. Furthermore, proper radiochemical preliminary operations were needed in order to evaluate the 90Sr specific activity [7,8]. An ultra-low background Perkin Elmer Quantulus liquid scintillator was employed as a Čerenkov beta counter to measure the 90Y specific activity. 3. RESULTS Preliminary results are reported in Table 1. Table 1. 137Cs and 90Sr specific activity [Bqg-1] in preliminary samples, where DL is the detection limit [9]. Sample 10 11 12 13 14 15 16 17 18 137Cs 1.9E-1± 2.5E-3 8.4E-2± 1.6E-3 1.2E-1± 2.1E-3 1.1E-1± 1.9E-3 2.1E-1± 2.7E-3 2.0E-1± 2.5E-3 1.1E-1± 1.9E-3 1.7E-1± 2.5E-3 1.8E-1± 2.5E-3 90Sr < DL < DL < DL < DL < DL < DL < DL < DL < DL 4. CONCLUSIONS The preliminary results show that no contamination was spread outside the reactor building. The radioactive levels of the contaminants are comparable with the average values of the region, which were considerably affected by the Chernobyl accident. REFERENCES [1]NRC, EPA, DoE, DoD, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), 2000. [2]CND (FP6), Environmental Radiation Survey and Site Execution Manual (EURSSEM), 2010. [3]EC, RP 122 - Recommendation on Practical use of concepts of clearance and exemption: Part I, 2000. [4]ISO 11932 - 1996. [5]ISO 18589-2 - 2007. [6]ISO 18589-3 - 2007. [7]ISO 18589-5 - 2009. [8]Eichrom Technologies, Inc., Strontium 89,90 in Water (SRW01), Analytical Procedures, 2003. [9]ISO 11929-1 - 2010.

Speakers: Mr Eros Mossini (Politecnico di Milano), Mr Luca Saverio Angelo Codispoti (Politecnico di Milano)

15:00 Geotechnical Application and Education Initiatives to Recovery of Nuclear Power Plant Accident

1. INTRODUCTION By off the Pacific coast of Tohoku earthquake in Japan on March, 2011, the subsequent accident of nuclear power plants in Fukushima has been problematic. Nuclear reactors have been damaged, and surrounding environment has been contaminated by radioactivity. The severe accident must be solved along with four tasks: demolition of damaged facilities, remediation of the site, waste management and preventing the spreading of contamination. So far, geotechnical engineering has technically contributed to radioactive waste disposal and general/industrial waste disposal. Sufficient results related to controlling groundwater and ground stabilization have been achieved in general construction projects. Both techniques and experiences of geotechnical engineering would be applicable to the tasks. The Japanese Geotechnical Society (JGS) has started a project that is being supported by the Ministry of Education, Culture, Sports, Science and Technology (MEXT). The project promotes both the technical development and the education of geotechnical engineering for decommissioning of nuclear power plants. This paper introduces contents of the JGS project. Geotechnical application and education initiatives to support the long term solution of nuclear power plant accident are discussed. 2. 2. RECENT SITUATION IN FUKUSHIMA AND GEOTECHNICAL VIEWPOINTS Factual details regarding the situation at the Fukushima I Nuclear Power Plant are still not clear, because the high levels of radiation around the site are preventing ground-based investigation. Currently, plans are being made to send in robotic samplers that can communicate electronically, and many nuclear, mechanical, and electronic engineers are now working on the development of such remote technologies. Outside the Fukushima I Nuclear Power Plant, remediation of contaminated ground is proceeding. Use, storage and disposal of the soil removed from the contaminated site are important practical issue at this moment. Geotechnical engineering has contributed to recycling by-products and waste, the environmental impact assessment in construction projects, construction and operation of waste disposal site, remediation of groundwater and soils contaminated by heavy metals, and planning geological disposal of radioactive waste. The situation caused by the accident at the Fukushima I Nuclear Power Plant would be improved to adopt the geo-environmental techniques. 3.RESEARCH AND EDUCATION PROJECT IN JAPAN As shown in Fig. 1, JGS focused on three subjects: (i) prediction for land and groundwater contamination by radionuclide from now until decommissioning, (ii) development of boring technology and excavation methods such as tunneling that can be used for excavation of the fuel debris remaining in the nuclear reactors, and (iii) planning and development for the disposal of large amounts of radioactive waste from the accident and decommissioning of the damaged nuclear power plant. Geotechnical engineering is able to make a variation of countermeasure against troubles in that site, which is a mind of fault tolerance in civil engineering. Research groups were systematically established in order to promote fundamental studies related to the subjects. As a preliminary investigation, Waseda University experimentally investigated a shielding technique against radiation using soils and the heavy slurry synthesized by metals and clays as shown in Fig. 2. Huge amounts of radioactive contaminated soils produced during the nuclear power plant accident are also big issue. According to the directive of the Ministry of the Environment in Japan, the excavated contaminated soils are temporally stocked, and it is gradually transported to the intermediate facility for the storage. A liner system of the facility, compaction of soils and designing cover soil are specialty of geotechnical engineering. Brand-new technologies are now developing to apply geotechnical engineering to the problem as soon as possible. Based on progress of the research, JGS has been working on an education program regarding fuel debris treatment and decommissioning of a nuclear power plant in Fukushima, supported by MEXT. Because it takes a long time to complete the remediation at Fukushima, it is necessary to educate younger persons who will become engineer and policymaker. “Radioactivity” is an advanced key word for geotechnical engineering and civil engineering, so the curriculum must be arranged by new viewpoints such as shielding and protection against radiation to be considered in conventional construction, operation and monitoring process. 4.CONCLUDING REMARKS Soil is composed by solid (soil particle), liquid (water) and gas (air) phases. The ratio of the three phases and density of soil can be artificially controlled. Therefore, soil is a useful natural material which has a shielding potential against radiation. To solve the issue caused by the accident at Fukushima I Nuclear Power Plant, geotechnical engineering must be developed still. The advanced geotechnical engineering will be established with both fundamental researches and in-situ education. JGS aims at cooperation with various academic fields, and networking to next generations, which are key toward the total solution of the severe accident such as a nuclear power plant in Fukushima.

Speaker: Dr Yasutaka Watanabe (Central Research Institute of Electric Power Industry)

15:00 Impact of radioactive waste management on dismantling cost

Abstract: It is important to estimate the cost of decommissioning at various stages of planning to manage decommissioning projects efficiently. Uncertainties regarding radioactive waste management including the availability of disposal facilities are making the estimate challenging. Some of the uncertainties such as unit disposal cost and transportation cost directly add up to the total decommissioning cost, while other factors may also influence the overall cost by limiting how the components and structures are dismantled in order i.e. for the waste to be segregated properly and fit into designated containers. In this study, impacts of radioactive waste management on dismantling cost was examined in order to plan decommissioning effectively while dealing with these uncertainties. Assuming a case of dismantling a reactor building of a 1100 MW BWR facility, labor cost and the amount of waste in each waste category were estimated for several cases with different waste container sizes. Other factors that may affect the dismantling cost are also examined including the timing when disposal facilities become available, i.e. duration of the storage of the dismantled waste before disposed. 1. INTRODUCTION A number of commercial nuclear power plants in Japan are approaching the end of their designed plant lives, and decommissioning of these facilities are expected to be major tasks in the next decades to come. It is necessary to estimate the cost of decommissioning at various stages of planning to manage these projects efficiently. Radioactive waste management cost, which constitutes a substantial portion of the total decommissioning cost, contains significant uncertainties including the timing when disposal facilities become available, unit disposal cost, distance from the decommissioning facilities to disposal facilities, mode of transport, and other specifics such as sizes and shapes of waste containers and the radiation dose limit of the waste container. Some of these factors such as unit disposal cost and transportation cost directly add up to the total decommissioning cost, while other factors may also influence the overall cost by limiting how the components and structures are dismantled in order i.e. for the waste to be segregated properly. In this study, impacts of radioactive waste management on dismantling cost was examined in order to plan decommissioning effectively while dealing with these uncertainties. 2. METHDOS Assuming a case of dismantling a reactor building of a 1100 MW BWR facility, labor cost and the amount of waste generated from dismantling were estimated using a similar method as in Smith et al. [1]. The low level waste generated from decommissioning is to be segregated into four categories according to the Japanese regulations; relatively high level (L1), low level (L2), very low level (L3) radioactive waste, and non-radioactive waste. Dismantling of reactor vessel and internals is a major process in decommissioning, and produces L1 and L2 level waste. Labor cost for cutting reactor vessel and internals was estimated to compare several cases that use different radioactive waste containers including 200 L drums, 400 L drums, and 1.6 m-cubed box. Reactor vessel and internals are first cut into a length packageable to containers, and then further cut up into smaller pieces so that each piece fits in the container. 3. RESULTS Total cutting length decreases when larger size containers are used, resulting in a decrease in the labor cost. Our estimate shows that by increasing the size of the L2 level waste from 200 L drums to 400 L drums, labor cost for cutting was estimated to decrease by approximately 30 %, and further increasing the size to 1.6 m cubed boxes reduced the estimated labor cost for cutting to approximately 50 % of the cost when 200 L drums were used. 4. CONCLUSIONS The result suggests that the size of the radioactive waste container is one of the factor that may influence the dismantling cost. Other factors that may affect the dismantling cost are also considered including the timing when disposal facilities become available, i.e. duration of the storage of the dismantled waste before disposed. REFERENCES [1] SMITH, R.I., et al., Revised Analyses of Decommissioning for the Reference Boiling Water Reactor Power Station, NUREG/CR-6174, (1996) .

Speaker: Dr Naoko Watanabe (Hokkaido University)

15:00 METHOD OF LIQUID RADIOACTIVE WASTE AND CONCENTRATES PURIFICATION FROM ORGANIC IMPURITIES BY MEANS OF ACTIVATED PYROLUSITE

The final stage of any method of liquid radioactive waste and concentrates conditioning is always recycled waste curing. Currently, the most commonly used methods of waste curing are: cementation, vitrification, and creation of ceramic-concrete matrices [1]. It is important to remove water from the cured product. However, it is impossible to completely remove water from the cement and ceramic matrices. Water contained in the concrete can be either in bound or free state. Free water contained in the concrete pores and interstices evaporates relatively quickly with temperature increase and concrete aging. As a result of hydrolysis of bound water, embrittlement of cement matrix occurs [2]. In case of a sealed container with a cement compound, gaseous products of radiolysis are accumulated and pressure increases. Cases of the container depressurization and destruction of the cement stone due to the excessive pressure of gaseous products (H2, H2O2) were observed [3]. Due to the damaging effects of radiolysis on cemented and vitrified waste, the new methods of waste solidification are currently developed. Russian researchers and their foreign colleagues are developing new ceramic-concrete matrices [4]. The use of magnesium-potassium phosphate matrix (MPP) requires maximum initial waste dewatering. Аs opposed to chemically bound water, which can be driven off to some extent by heating of the compound, the organic impurities in the waste cannot be distilled this way. Therefore, their preliminary chemical decomposition is necessary. Alternatively to the above methods, the use of activated pyrolusite as inorganic sorbent is considered in this paper. Activation means creation of Mn6+ surface layer from MnO2. Method proposed by the authors is the sorption of organic impurities from the liquid radioactive concentrate by activated pyrolusite followed by evaporation of the resulting slurry to dryness. The studies showed that soaps (sodium oleates, stearate, and palmitates) were most effectively adsorbed by activated pyrolusite from the anionic surfactants. It should be noted that in the experiments with a series of detergent solutions, catalytic decomposition of anionic surfactants by activated pyrolusite to oxides constituting the organic molecule elements, namely: SO2, CO2, and H2O, was recorded [5]. The following conclusions can be drawn from the obtained data: -method of liquid radioactive waste and concentrates purification from organic impurities by means of activated pyrolusite was proposed; -the method is applicable for the immobilization of liquid radioactive waste and concentrates with 60 - 150 g / L or higher concentration of organic impurities; -when heated, activated pyrolusite acts as a catalyst for decomposition of organic impurities to the oxides of the constituent elements (CO2, SO2, and H2O). In this regard, sorbent separation from the solution is not needed, this greatly simplifying the process of liquid radioactive waste purification; -the method doesn`t require to include into the flow sheet filtering materials and components that need further processing; -the use of the activated pyrolusite allows to immobilize waste sorbent with salt slurry in any matrix for the long-term storage of solidified radioactive waste.

Speaker: Ms Christina Legkikh (JSC “SSC RF – IPPE”)

15:00 PREPARATORY WORK FOR THE DECOMMISSIONING OF RBMK-1000 POWER UNITS OF LENINGRAD NPP

The decommissioning of a nuclear power unit, as any radiation-hazardous facility, is an integral and inevitable stage of a nuclear facility life cycle. NPP decommissioning implies sequence of administrative and technical activities intended to terminate any activity connected with the facility functional purpose, and make it environmentally benign to be released from regulatory control[1]. We are nearing the life extension of Leningrad NPP power units, with the shutdown of Unit 1 scheduled for the end of 2018. The following decommissioning documentation has been developed at Leningrad NPP: - Concept of Leningrad NPP power units decommissioning[4]; - Decommissioning Program[2]; - Program of Integrated Engineering and Radiation Survey (IERS) for Units 1 & 2; - Safety Case for Units 1 & 2 shut down for decommissioning[3]; - Detailed documentation requirements specification for Units 1 & 2 decommissioning; The following documents are about to be issued: - TechSpecs for the period of decommissioning; - Design of Units 1 & 2 decommissioning. Leningrad NPP developed the program specifying arrangements for the management of failed/damaged spent nuclear fuel assemblies and spent elements of RBMK-1000 reactor cores. Underway are activities connected with the implementation of dry storage of spent nuclear fuel and its transportation to Mining and Chemical Combine (Zheleznogorsk). In progress is the preparation of documentation pertaining to spent nuclear fuel afterburning at operating power units of Leningrad NPP.

Speaker: Mr Pavel Gredasov (Open Joint Stock Company «Concern for Production of Electric and Thermal Energy at Nuclear Power Plants» («Rosenergoatom» Concern OJSC) Leningrad NPP)

15:00 THE OPERATION OF NON-GOVERNMENTAL ORGANIZATIONS IN THE FIELD OF ECOLOGICAL CONTROL OF DECOMMISSIONING NUCLEAR SITES

Nuclear decommissioning is the process whereby a nuclear power plant site is dismantled to the point that it no longer requires measures for radiation protection. The presence of radioactive material necessitates processes that are occupationally dangerous, hazardous to the natural environment, expensive, and time-intensive.[1] Decommissioning is an administrative and technical process. It includes clean-up of radioactive materials and progressive demolition of the plant. Once a facility is fully decommissioned, no radiologic danger should persist. The costs of decommissioning are spread over the lifetime of a facility and saved in a decommissioning fund. After a facility has been completely decommissioned, it is released from regulatory control and the plant licensee is no longer responsible for its nuclear safety. Decommissioning may proceed all the way to "greenfield" status [2]. The life cycle of the NPP comprises steps of: design, construction, commissioning, operation and maintenance, nuclear waste management, development, decommissioning. An important step for the social society is the decommissioning. In this case, there is deterioration in the economic and environmental climate in the region where the NPP was operated. Perhaps an important role can be played by environmental NGOs. The goals of environmental NGOs include but are not limited to: creating relationships with the government and other organizations, offering training and assistance in agricultural conservation to maximize the use of local resources, establishing environmental solutions, and managing projects implemented to address issues affecting a particular area [3]. Environmental NGOs are organizations that are not run by federal or state governments but rather have funds issued to them by governments, private donors, corporations, and other institutions [4]. In order to fully understand the social, economic, and environmental effects an organization can have on a region, it is important to note that the organization can act outside the formal processes that state governments and other government institutions must comply with. It was decided to develop a program of «Non-governmental monitoring in nuclear power» at International Youth Forum of Power Engineers and Industrialists «Forsage 2015». Nuclear Experts from CIS countries decided to invite environmental NGOs to jointly develop such a program. There are some environmental NGOs in Russian Federation. One such organization is the "OKA" NGO. The organization has experience in environmental expeditions to nuclear power plants and other nuclear sites. The development of this program for environmental NGOs will increase the effectiveness of public environmental monitoring and the radiation safety of nuclear facilities. It will also improve a public opinion towards nuclear energy. References [1] Benjamin K. Sovacool. "A Critical Evaluation of Nuclear Power and Renewable Electricity in Asia", Journal of Contemporary Asia, Vol. 40, No. 3, August 2010, p. 373. [2] "Fact Sheets: Decommissioning Of Nuclear Power Plants". National Energy Institute. Retrieved 2014-06-19 [3] Keese, James R. (1998). "International NGOs and Land Use Change in Southern Highland Region of Ecuador". Human Ecology 26: 451–468. [4] Pamela Chasek, ed. 2000 The Global Environment in the Twenty-First Century: Prospects for International Cooperation. United Nations University.

Speaker: Dr Pavel Belousov (INPE NRNU MEPhI)

15:00 The repository for the radioactive wastes class 3 and 4 basic design. JOI.

Nuclear cycle facilities in Russian Federation by now has accumulated (and continue accumulating) a large amount of radioactive wastes, which contains different subjects with the complex morphology and various nuclide set [1, 2]. The only way to solve the problem of the long-term radioactive wastes storage and complying all the up-to-date safety requirements is to create the modern repository as a capital construction. There are no repositories of the sufficient capacity and environmental stability created by ROSATOM being operated yet. In fact, it means that all the design work and engineering surveys have still to be executed. The article discusses the basic design of the repository for the radioactive wastes class 3 and 4 in terms of justification of investments (JOI) is going to be realized at the Production Association «MAYAK». The variants of this basic design project providing the economic efficiency of the radioactive wastes management, based on the repository plant geological and ecological features [3-5] are considered. Such a repository concept is proposed to perform two main functions: 1) safe storage of the solid radioactive wastes of low and medium volume activity with industrial quantities provided by the utility systems (passive or active); 2) exception of the radionuclide migration through the protecting barriers provided by the localization constructions of the repository building. List of references 1. Guidance on radioactive waste management legislation for application to users of radioactive materials in medicine, research and industry. IAEA-TECDOC-644, Vienna, 1992. 2. Radioactive Waste Management: Status and Trends – Issue #3. IAEA/WMDB/ST/3. August 2003. 3. Technical, institutional and Economic Factors Important for Developing a Multinational Radioactive Waste Repository. IAEA-TECDOC-1021. Vienna. 1998. 4. Regional and International Solutions for Long-Lived Radioactive Waste Disposal: the ARIUS initiative // Proceedings of the 10th International IHLRWM conference in Las Vegas, Nevada, March 30 - April 3, 2003. 5. ElBaradei, M. Geological Repositories: The Last Nuclear Frontier. // International Conference on Geological Repositories. 8-10.12.2003, Stockholm, Sweden. www.iaea.org.

Speaker: Mr Andrei Gordeev (JSC "CPTI")

16:30 → 17:00 Coffee break

16:35 → 17:00 Coffee break

17:00 → 18:00 Session 6 - 2: Project Management, Skills, and Supply Chain Considerations

The purpose of this session is to highlight how experience over the past 10 years has impacted the supply chain and identify future trends – to highlight current good practice in terms of:
• development and implementation of programmes and projects; and
• highlighting needs, opportunities and challenges related to knowledge management, education and training to support future project implementation

17:00 → 18:00 Young Professional Session - 2

Friday, 27 May

09:00 → 11:00 Session 7: Future Needs and International Cooperation

Focus on financial contribution, infrastructure requirements, but also knowledge transfer – highlight barriers and mechanisms to overcome

11:00 → 11:30 Coffee break

11:30 → 13:00 Conference Closing Session