International Conference on Research Reactors: Safe Management and Effective Utilization

16 Nov 2015, 08:00 → 20 Nov 2015, 19:00 Europe/Vienna

VIC Board Room A

Research Reactors: Safe Management and Effective Utilization

The purpose of this conference is to foster the exchange of information on operating and planned research reactors and to provide a forum at which reactor operators, managers, users, regulators, designers and suppliers can share experience and lessons learned, as well as address common issues, challenges, and strategies.

Monday, 16 November

08:00 → 12:00 Registration

Registration of conference participants

09:30 → 10:30 Official Opening

Official Opening of the Conference

10:30 → 11:00 Coffee Break

Coffee Break

11:00 → 12:30 Utilization and Application: I

Convener: Prof. Helmuth Boeck (Vienna University of Technology/Atominstitut)

11:00 IAEA Activities in Support of Enhanced Research Reactor Utilization and Sustainability

The underutilization of research reactors (RRs) around the world persists as one of the primary concerns to global nuclear research and technology development, and threatens the sustainable operation of individual RRs. The IAEA responds with a broad range of activities to address the strategic planning, execution, and utilization improvement of many of these facilities. First of all, the revision of two critical documents for enlarging and diversifying a facility’s portfolio of applications has been undertaken in order to keep this information up-to-date, corresponding to the dynamism of experimental techniques and research capabilities, including services and products RRs can provide. In 2014 the release of Nuclear Energy Series NP-T-5.3, “Applications of RRs” has replaced the well-known TECDOC 1234 (2001). The review process of the TECDOC 1212 “Strategic Planning for RRs” (2001) has been finalised, with a revised and enlarged publication expected in 2015. Secondly, the IAEA continued supporting regional RR networks and coalitions, what helps foster the cooperation and assists facilities in expanding their stakeholder base and user community. Presently, the initiative accounts for 8 sub-regional entities created with more than 50 member states involved (including over 20 countries without such facilities). Thirdly, a number of thematic networking activities focusing on specific applications of RRs are either continuing or starting, with a few selected examples listed below: • In order to reflect the current status and trends in RR utilization and management, a group of international experts reviewed 37 strategic plan documents submitted by managers around the world. As a follow up to the review, two interregional workshops were organized in 2013 and 2014, which gave the opportunity to RR managers to share their experiences, lessons learned and good practices in developing and implementing strategic plans at their facilities. • In neutron activation analysis (NAA), an extensive round of inter-comparison proficiency testing was completed (2010-2013), resulting in identification of poorly performing laboratories and implementation of necessary corrective measures through the follow up workshops, procurement of reference materials and dedicated expert review missions. Two new inter-comparison rounds are taking place in 2015. Additional support is ensured through an active CRP on “Development of an Integrated Approach to Routine Automation of NAA” (2012-2015). • The first attempt towards standardization of digital neutron imaging was successfully completed in 2013 thanks to the Round Robin exercise organized and supported by the IAEA. Further improvement of the reference samples and measurement protocols were discussed in the follow up technical meetings. A dedicated CRP is planned in 2016. In addition, a series of training workshops are being organized with hands-on-training at the well-established neutron imaging facilities (in 2013 and 2015). • In the area of the education and training, IAEA already supported 10 iterations of a six-week group fellowship training course on RR safety, utilization, O&M, hosted by RRs of the Eastern European Research Reactor Initiative (EERRI) and resulting in close to 80 young professionals trained since 2009. The Internet Reactor Laboratory project is another important initiate which presently is being extended to other regions after a successful pilot demonstration between NCSU/USA and JUST/Jordan. • Other support and assistance are provided through close to 30 national and regional Technical Cooperation projects, all, among other areas, relevant to the applications and enhanced utilization of RRs. This presentation will briefly describe the above efforts and introduce future activities in the area RR networking and regional/international cooperation.

Speaker: Dr Danas Ridikas (IAEA)

11:30 THE IMPACT OF THE EDUCATION AND TRAINING ON RESEARCH REACTORS FOR THE BUILDING OF THE KNOWLEDGE, COMPETENCIES AND SKILLS

The development or the sustainability of the nuclear programs needs the availability of human resources with adequate knowledge, competencies and skills. In the past, countries that have embarked on nuclear program went first into the design, construction and/or utilisation of research reactors as a first step in the development of the human capacity for their program. Nowadays, some countries are considering a slightly different approach, embarking on a nuclear program without having a research reactor of their own. Nevertheless, it appears that their human capacity building relies on the use of knowledge, competencies and skills of experienced staff that benefited from the utilisation of research reactors. This state of the art tends to show the importance of the research reactors for the development of the human capacity for the nuclear programs. On the other hand, with the fast development of the technology and in particular of the software applications, calculation codes and simulators (including full scale simulators), it may be considered that such tools could fully replace the utilisation of research reactor for the development of the knowledge, competencies and skills of the personnel involved in reactor design, construction, operation and safety. This subject is a question of debate between those who think that simulators can replace the practical experience on research reactors and those who think that this practical experience cannot be fully replaced by the simulators. This paper is concerned with the study of the knowledge, competencies and skills that are developed on the research reactors and on the simulators. It compares the contributions of these two different tools to the building of the expertise. It identifies what are the learning outcomes specific to one of these tools and what are those common to both of them. The paper concludes that these two tools are not competing but that they are complementary. In fact, each tool brings some specific bricks in the development of the knowledge, competencies and skills that cannot be gained from the other one. Further analysis emphasises the importance of the practical experience on research reactors to maintain or to develop the expertise of the personnel involved in the nuclear programs. It shows the impact of this practical experience in making the link between the general principles (in various fields such as reactor principles, neutron kinetics, safety…) or the calculations (core design and configuration) with the real behaviour and operational constraints of a reactor. Finally, it shows how important is this practical approach for the development of the safety culture amongst all the personnel involved in the safe operation and utilisation of the nuclear reactors.

Speaker: Mr François FOULON (CEA / INSTN)

11:50 A Comparative Analysis of the Use of Internet Reactor Laboratory and a Subcritical Assembly for Nuclear Engineering Education

Laboratory education is an essential component in any nuclear engineering education program, where research reactors play an important role. However, having a research reactor solely for educational purposes is faced with many challenges such as the high cost, long construction time, running cost, safety, and security associated issues. In 2007, Jordan University of Science and Technology (JUST) established the first and only Nuclear Engineering program in Jordan with the objective to provide Jordan’s nuclear energy program with qualified nuclear engineers and support its human capacity building program. JUST offers a Bachelor of Science (BSc) of 160 credit hours in Nuclear Engineering over a period of five years. One of the compulsory courses is a Nuclear Reactor Laboratory Course. For the years 2010-2013 the course was taught using the Internet Reactor Laboratory utilizing the PULSTAR reactor at North Carolina State University. From 2014 until today the course is being taught using Jordan Subcritical Assembly (JSA), which was constructed at the Department of Nuclear Engineering at JUST for the purpose of education, training, and experimental research. The work presented in this paper discusses the utilization of JSA for educational purposes as compared to the Internet Reactor Laboratory approach. Emphasis will be laid on meeting the learning objectives and outcomes of the nuclear reactor laboratory course.

Speaker: Dr Salaheddin Malkawi (Jordan University of Science and Technology)

Slides A03_Malkawi.pdf

12:10 Education and Training at Research Reactors: Sharing the Experiences from Europe with Asia and Africa

More than 770 research reactors have been built and operated since 2 December 1942 when the first reactor Chicago pile-1 reached first criticality. Nowadays approximately 250 research reactors in 56 countries are still in operation. Research reactors together with power generating reactors (power reactors) are the most common nuclear installations in the world. A research reactor can be used in a wide range of nuclear activities, from supporting nuclear technology (nuclear power generation) through nuclear science, neutron activation analysis, radioisotope production, neutron imaging up to and education and training. Research reactors are most widely used for education and training, according IAEA research reactor database 164 reactors from 247 in operation declare that a reactor is used for education and training. This trend began at the early stage of development of research reactor in mid-fifties of the last century when the first research reactors were built and many of them were specifically designed for education and training, for example for education of first generations of students in reactor physics (the first ARGONAUT reactor, February 1957) or to train the first generations of nuclear engineers in the USA (the first AGN-201, January 1957) as well as the TRIGA reactors (the first reactor in May 1957) where T in the reactor name means training. Research reactors are excellent tools for experimental education and training and hands-on activities in wide range of nuclear activities, such reactor physics, nuclear safety, radiation protection, neutron application or neutron science. Education and training at research reactors has long tradition in Europe where it is several decades regular part of educational processes at universities in France, Czech Republic, Austria, Slovenia, Hungary, Germany, Italy or Finland. Several European research reactors have bilateral agreement of collaboration in nuclear education and training with research reactors and Africa and Asia, but the most collaboration is carried out under various IAEA activities and projects. Four-years IAEA TC project RAF/4/022 (2009 to 2013) and two years IAEA TC project RAF/1/1005 (2014 to 2015) both focused on African research reactors and its users from Africa, allowed carrying out several activities in nuclear education and training where experiences from European research reactors were successfully used. Based on experiences from EERRI – Eastern European Research Reactor Initiative dedicated IAEA questionnaire has been developed where the capabilities and needs in nuclear education and training in Africa are described in details. One regional workshop and one regional training course were also organised with strong involvement of lecturers from European research reactors. Regional workshop on enhanced use of research reactors for education and training purposes was organised in Rabat at Moroccan TRIGA reactor in 2013 which was attended by 12 participants from 7 African countries (Algeria, Egypt, Ghana, Libya, Nigeria, Sudan and Tunisia). Regional training course on the safety of research reactors was also organised in Rabat in 2014 which was attended by 12 participants from 7 African countries (DRC, Egypt, Ghana, Libya, Nigeria, South Africa and Sudan). Another ongoing project where European research reactors are involve in Africa is Internet reactor laboratory where French reactor ISIS will serve as host reactor for broadcasting of education to several African countries. Similarly as in Africa, several European research reactors established bilateral collaboration with research reactors in Asia. During IAEA regional workshop on education and training practices with research reactors organised in 2012 in Prague the first contacts between European research reactor and research reactors in South - East Asia have been established. Under IAEA PUI project the first regional training course with strong involvement of lecturers from European research reactors was organised in Selangor at Malaysian TRIGA reactor in 2013. This training course on education and training to support nuclear power program in APEC economies attended 30 participants from Malaysia and Thailand. The second South - East Asia regional course was named “Nuclear School Experiments on Reactor Physics and Neutron Applications for Asia-Pacific Region”. In this course for the first time two research reactors from two countries were involved. The course was organised in 2015 in Selangor Selangor at Malaysian RTP reactor and in Jogyakarta at Indonesian KARTINI reactor. The nuclear school was attended by 11 participants from 7 Asian countries (Bangladesh, Cambodia, Indonesia, Iraq, Malaysia, Thailand and Vietnam). European experiences also have been used during development of reactor experiments at Malaysian RTP reactor for students from Malaysian universities.

Speakers: Mr François FOULON (CEA / INSTN), Dr Lubomir Sklenka (Czech Technical University)

Slides A04_Sklenka.pdf

12:30 → 14:00 Lunch Break

14:00 → 15:30 Utilization and Application: Utilization and Application II

Convener: Dr Peter BODE (Delft University of Technology)

14:00 Neutron imaging in science and technology

Neutron imaging is a method of non-destructive investigation for objects of scientific and technological interest. Within the last decade, neutron tomography and radiography have significantly gained importance among the neutron science community. One of the reasons is the fast development in digital image recording and processing technology, which has allowed overcoming of some previous limitations in spatial and time resolution. Another reason is that in addition to the attenuation contrast technique, new innovative methods for neutron imaging are being implemented. Using monochromatic neutrons for imaging, complementary contrast due to the coherent scattering in polycrystalline materials can be obtained, providing information about structural changes or composition inhomogeneities. In addition, neutrons possess a magnetic moment which makes them sensitive to magnetic fields. The magnetic interaction can be used for two- and three-dimensional visualization of magnetic field distributions both in free space and in bulk materials. Utilization of phase contrast and dark-field contrast techniques e.g. using grating interferometry allows for visualization of low-absorbing materials and microstructural heterogeneities, as well as magnetic structures (domain walls). The neutron tomography instrument CONRAD-2 had been in operation since 2005 at the Hahn-Meitner research reactor at Helmholtz-Zentrum Berlin (HZB). Over the last years, significant development work has been performed to expand the radiographic and tomographic capabilities of the beamline. New techniques have been provided to the user community as tools to help address scientific problems over a broad range of topics such as superconductivity, materials research, life sciences, cultural heritage and paleontology. Industrial applications including fuel cell research have also been improved through these new developments. The instrument is used for educational purposes in the frame of the annual Berlin School on Neutron Scattering. Various studies related to a large number of diploma and doctoral theses have been performed at the beamline.

Speaker: Dr Nikolay Kardjilov (Helmholtz-Zentrum-Berlin)

Slides A05_Kardjilov.pptx

14:30 The Use of Thermal Neutron Beams at Medium Power Reactor LWR-15 in Řež for Competetive Neutron Scattering Experiments

Neutron Physics Laboratory of NPI ASCR, v.v.i. operates several neutron instruments installed at the medium power reactor LWR-15 having a nominal thermal power of 10 MW and a neutron flux rate of about 1x1014 n.cm-2s-1 in the core. Research investigations at an international level are carried out using dedicated instruments. They are • Determination of internal stresses in polycrystalline materials by neutron diffraction. This instrument is used especially for thermo-mechanical testing of materials, i.e. to study the deformation and transformation mechanisms of modern types of newly developed materials. Then, neutron diffraction performed in situ upon external loads, brings a wide range of valuable structural and sub-structural parameters of the studied material which can be easily correlated with parameters of the external loads. • Microstructure, porosity and inhomogeneity studies by a high Q-resolution small-angle neutron scattering (SANS). The SANS instrument is mainly suited for investigation of structural or compositional inhomogeneities in materials in the size range 0.05÷2 micrometer, mainly porous materials and large precipitates in alloys. • Structure studies of polycrystalline materials by medium resolution powder diffraction and structure behaviour under termo-mechanical load. The powder diffractometer is mainly used to the study of the crystalline and/or magnetic structure of the powder or polycrystalline samples. Due to the several sample environments including vacuum and light furnace, close cycle cryostat, Euler goniometer, deformation rig and automatic sample exchanger, this instrument is an universal tool in the field of powder diffractometry. • Surface studies of technologically interesting materials by means of neutron depth profiling (NDP). The NDP technique exploits nuclear reactions of neutrons with nuclei to analyze concentrations or concentration profiles of elements in solids just under the surface. • Neutron activation analysis (NAA) using vertical irradiation channels. Using both short-time (10 s - 3 min.) and long-time (several hours - several days) irradiations, information about concentrations of up to 65 elements can be obtained, in many cases by non-destructive, so-called instrumental neutron activation analysis (INAA). The detection limits range from μg.kg-1 up to tens of percent, depending on the particular element and bulk matrix composition. • The development of high resolution and high efficient monochromators and analysers on the basis of Bragg diffraction optics using cylindrically bent perfect crystals. • Education and training programmes. NPI has also capability for Monte Carlo simulations by using a widely recognized (at the European level) RESTRAX software package for estimation of characteristic properties of designed performance of neutron scattering devices and their optimization. The unique of the RESTRAX package for neutron ray tracing consists in its possibility of implementing the neutron optical elements (e.g. curved crystals, flat or curved mirrors and supermirrors, any type of collimators etc.). In all the investigations, highly recognized results of an international significance have been obtained during the last decade. Recently, substantial work has been done on the Engineering diffractometer BEER which will be installed at ESS neutron source. A high standard of neutron research in the Neutron Physics Laboratory of NPI has been recognized by participation of the laboratory in EU-NMI3 ACCESS projects as well as in the next one EU-SINE2020 starting in 2016. The access to the mentioned experimental techniques for external users is for the future ensured by a Czech project CANAM (http://canam.ujf.cas.cz/index.php?option=com_flexicontent&view=items&id=142&lang=en). Examples of some unique experimental results obtained on the NPI instruments installed at the reactor LWR-15 will be presented.

Speaker: Dr Pavol Mikula (Nuclear Physics Institute ASCR, v.v.i., 250 68 Rez, Czech Republic)

Paper Mikula_IAEA_Paper.pdf

Slides 37_IAEA_conference_presentation.pdf

14:50 Modern methods to test materials and fuel in research reactors of RIAR

RIAR uses the SМ-3, MIR.М1, BOR-60, RBT-6 research reactors to test materials and fuel. By their design and purpose, all the RIAR reactors refer to test reactors and are used for applied research in the field of nuclear power engineering and radioisotopes production. The innovative nuclear power systems of the next generation must be more efficient and safe. To solve these tasks there are necessary the new materials with higher properties, i.e. to carry out research activities, investigations and examination of properties of different materials under irradiation. It should be mentioned that RIAR’s research reactors have comprehensive and harmonic capabilities to test fuels and materials for all existing and innovative projects of water-cooled reactors, including super-critical water parameters with moderated neutron spectrum, and fast reactor with liquid metal and gaseous coolants.

Speaker: Mr Alexey Izhutov (JSC “SSC RIAR”)

15:10 Overview of the NEA Activities Related to Experimental Needs

The Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) performs works on experimental needs identification and prioritisation for potential applications including the safety and design of nuclear installations. Thousands of experiments conducted in the past have been collected, documented and preserved to support benchmarking, general reactor physics and material science studies. Most of these data are stored in the relevant databases by the NEA Nuclear Science Division (NSD) and Data Bank teams. A major source of data used for code validation in criticality safety and neutron physics are the NEA Handbooks of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP). They provide the nuclear community with peer reviewed benchmarks that include critical, subcritical, shielding, radiation-transport, fundamental physics and reactor physics experimental data from nuclear facilities, worldwide. Selected data from each experimental configuration is entered into the corresponding databases: Database for the International handbook of evaluated Criticality safety benchmark Experiments (DICE) and the IRPhEP Database and Analysis Tool (IDAT). Research reactors have been an important source of information for the Projects. They contain evaluated cases of the experiments performed on the TRIGA type reactors, the Advanced Test Reactor (ATR), USA, the High Temperature Engineering Test Reactor (HTTR) and JOYO, Japan, the HTR-10, Chinese small pebble-bed test reactor and others. More than 70 shielding benchmark experiments are collected in the SINBAD database that is jointly developed and maintained by the OECD/NEA Data Bank and Oak Ridge National Laboratory, Radiation Safety Information Computational Center (ORNL/RSICC). Many of these benchmarks were carried out at research reactors to validate shielding calculations, and are particularly useful for advanced shielding designs for new research or power reactors. Research reactors play an important role in providing data for safety assessments. Several experimental projects are conducted on the research reactors under the auspices of the NEA Committee on the Safety of Nuclear Installations (CSNI). Major ones among them are the Halden Reactor in Norway, the Cabri reactor in France, the Japanese HTTR and the Loss-of-Fluid-Test (LOFT) reactor in the USA. Work on identification of experimental needs for safety of nuclear installations has been performed by the CSNI. Thus, a collection of internationally agreed matrices of experimental benchmarks derived from the separate and integral effect tests was developed for validation of best estimate thermal-hydraulic computer codes. These data support the validation process in order to increase confidence in the predictive capability of codes for existing and advanced systems, including the quantification of the uncertainty range for the simulation models and methods. The CSNI working groups issued a series of state-of-the-art reports on fuel behaviour in various accident conditions where experimental needs were considered mainly for LWR requirements. The NSD activity on fuel behaviour - the Thermodynamics of Advanced Fuels - International Database Project - was established in 2013 to make available a comprehensive, internationally recognised and quality-assured database of phase diagrams and thermodynamic properties of advanced nuclear fuels. With the new trends in nuclear power generation and regulation, design of nuclear installations, and experimental capabilities, the need for new integral experiments at large scale remains a high priority and requires a cross-disciplinary approach. Given these conditions, the periodic revision of experimental needs and available measurement capabilities are necessary in order to address the requests of industry, safety assessment, and the scientific community. In this context, a new NSD activity has been focused on identification of experimental needs for neutronics, thermal-hydraulics, material studies, fuel behaviour, and multi-physics. The anticipated output from this activity will be establishment of a framework that will bring together international experts in order to revise the experimental needs, rank priorities for validation and identify experimental facilities where the needs could be addressed. The development of advanced simulation methods, particularly coupled multi-physics methods, has been a significant trend in recent years. Along with the progress made in computational capabilities, new requirements for integral data arise in order to meet the validation process demands. The operational flexibility of most research reactors allows them to address the major needs identified for the nuclear industry providing testing and calibration experiments, integral experiments, benchmarking, code validation analyses, and cross-section measurements. The NEA Research and Test Facilities Database (RTFDB) that contains description of about 700 experimental facilities, including research reactors was created in 2007. Recently, work has been started on database modernisation that is focused on checking and extending the data collection as well as providing users with easy access to the parameters of experimental facilities and links to the information available in existing NEA databases. The full paper will describe progress the revision of experimental needs. It will provide a current status of the NEA databases and the projects related to safety assessment.

Speaker: Dr Tatiana Ivanova (OECD Nuclear Energy Agency)

Paper IAEA_Paper_121(FIN).docx IAEA_Paper_121(FIN).pdf

Slides A08_Ivanova.ppt

15:30 → 16:30 Break: Poster Session and Coffee Break

16:30 → 18:10 Common Management Considerations

Convener: Prof. Natesan Ramamoorthy (Bhabha Atomic Research Centre)

16:30 Management of the IAEA Cross Cutting Activities on Research Reactors

For several decades, research reactors have been a corner stone in the development and application of nuclear science and technology and in education and training of nuclear scientists and engineers. The benefits of research reactors also extended to many other fields including medical and industrial applications. Research reactors also can play an important role for building the national infrastructure in countries embarking on nuclear power programmes. Continued safe and secure operation, and effective utilization of research reactor facilities is essential for the benefits of the whole nuclear community. Through its programmes and activities, the IAEA continues to provide a main contribution to enhancing safety, security, and effective operation and utilization of research reactors worldwide. These programmes and associated activities are continuously adapted to address the needs of Member States, issues and trends, and challenges facing the research reactor community. These programmes are established with the objectives to enhancing safety and security, improving utilization, promoting research and technological developments, supporting research reactor fuel cycle activities, supporting new research reactor projects, and minimizing the use of high enriched uranium (HEU). The IAEA plays a leading role in coordinating the worldwide efforts in these areas and in supporting Member States to address the relevant issues and challenges. These issues and challenges include regulatory effectiveness, safety and security management, ageing of reactor facilities and personnel, lack of adequate utilization, increased pressure for increased vigilance with respect to non-proliferation, and the need for establishing an adequate national infrastructure supporting the development of new research reactor programmes. These compromise a wide range of elements in safety, operation, fuel cycle, and utilization areas, which require effective coordination and harmonization of methods and approaches. The IAEA activities on research reactors are managed by several organizational units within the Department of Nuclear Sciences and Application (NA), Department of Nuclear Energy (NE), and Department of Nuclear Safety and Security (NS). Utilization and application activities for research reactors are led by NA and mainly oriented to supporting Member States in assessing their needs for research reactors and in improving effective utilization of the facilities for development in nuclear science and application. The technological aspects related to research reactor design, operation, maintenance, fuel cycle, HEU use minimization are managed by the NE which also coordinates the activities related to development of the national infrastructure to support new research reactor projects. NS is supporting Member States to build capacity in all aspects related to nuclear safety and nuclear security in all stages of the research reactor lifetime, including siting, design, commissioning, operation, utilization, and decommissioning. Additionally, the IAEA Department of Technical Cooperation supports research reactor activities in the TC recipient countries. NA, NE and NS support TC in the technical implementation of the TC projects and programme on research reactors. To ensure optimization of resources and harmonized approaches for the benefits of Member States, an IAEA Cross-cutting Coordination Group on Research Reactors (CCCGRR) has been established with representatives from all IAEA organizational units have activities on research reactors. The role of the Coordinator of the CCCGRR is assigned to NA, NE, and NS on rotational basis. After a brief review of the status of the research reactors worldwide and presentation of common issues and trends of a cross-cutting nature among nuclear safety, nuclear security, operation and utilization, the paper presents, along with discussions the ongoing and planned IAEA activities addressing these issues in a coordinated manner. These include, among others, activities on the development of national infrastructure for countries planning to build their first research reactor, managing ageing research reactors, conduct of coordinated research projects, conversion of research reactor fuel from highly to low enriched uranium, and fostering networking for safety and utilization improvements, and development of human resources. These activities have an important role in supporting Member States in effective application of the IAEA Code of Conduct on the Safety of Research Reactors, safety standards, and security guidance documents.

Speaker: Mr Amgad Shokr (IAEA)

17:00 Integrated management system, configuration and document control for research reactors

An integrated management system is a single management framework establishing all the processes necessary for the organisation to address all its goals and objectives. Very often only quality, environment and health & safety goals are included when referred to an integrated management system. However, within the research reactor environment such system should include goals pertinent to economic, environmental, health, operational, quality, safeguards, safety, security, and social considerations. One of the important objectives of an integrated management is to ceate the environment for a healthy safety culture. Configuration management is a disciplined process that involves both management and technical direction to establish and document the design requirements and the physical configuration of the research reactor and to ensure that they remain consistent with each other and the documentation. Configuration is the combination of the physical, functional, and operational characteristics of the structures, systems, and components (SSCs) or parts of the research reactor, operation, or activity. The basic objectives and general principles of configuration management are the same for all research reactors. The objectives of configuration management are to: a) Establish consistency among design requirements, physical configuration, and documentation (including analyses, drawings, and procedures) for the research reactor; b) Maintain this consistency throughout the life of the research reactor, particularly as changes are being made; and c) Retain confidence in the safety of the research reactor. The key elements needed to manage the configuration of research reactors are design requirements, work control, change control, document control, and configuration management assessments. The objective of document control is to ensure that only the most recently approved versions of documents are used in the process of operating, maintaining, and modifying the research reactor. Document control helps ensure that: 1. Important facility documents are properly stored; 2. Revisions to documents are controlled, tracked, and completed in a timely manner; 3. Revised documents are formally distributed to designated users; and 4. Information concerning pending revisions is made available. Configuration and document management within an integrated management system are essential requirements for the safe operation, utilisation and modification of any research reactor.

Speaker: Mr Koos Du Bruyn (NECSA)

Paper 151_Integrated_Management_System_Configuration_and_Document_Control_for_Research_Reactors.docx

Slides B02_du_Bruyn.pptx

17:30 World of TRIGA Research Reactors: Present and Future

TRIGA reactors (Training, Research, Isotopes, General Atomics) constitute a ‘World of their Own’ among the large variety of research reactors. Developed in the 1950s, they were mainly constructed in the 1960s and 1970s all over the world and many of them continue to operate successfully more than 50 years. In 1970, the first of many TRIGA Users’ Conferences was organised by the TRIGA reactor in Helsinki/Finland, followed by US- and European TRIGA Users’ Conferences until 2000. In 2002 the first combined World TRIGA Users’ Conference took place in Pavia/Italy, later on these conferences were added as a topical event to the annual Research Reactor Fuel Management Conferences (RRFM) organised by the European Nuclear Society, and to the annual conference of the Training, Research and Test Reactor (TRTR) organization in the United States. Majority of these Conferences are documented in a dedicated CD-ROM compiled by the AtomInstitut in Vienna together with the IAEA INIS Section [1]. This publication collects extensive volume of papers and presentations from the TRIGA conferences from 1970 to 2008; it includes more than 1000 searchable contributions from individual TRIGA facilities worldwide. As both the front end and the back end of TRIGA fuel became increasingly important to TRIGA reactors, the IAEA took the initiative to invite all TRIGA reactor operators to a dedicated Technical Meeting in Vienna/Austria in November 2013 to discuss important TRIGA issues and challenges, such as utilization, management, technical support, fresh fuel supply and spent fuel options. One of the outcomes of this meeting was the unanimous decision to compile a TECDOC, describing history, present status and future perspectives of TRIGA facilities worldwide. In April 2015 the TECDOC reached his final review stage. It covers both the Historical Development (Chapter 2) and basic TRIGA Characteristics (Chapters 2 and 3), followed by TRIGA Utilization (Chapter 4), TRIGA Fuel Conversion (Chapter 5) and Ageing Management of TRIGA research reactors (Chapter 6). The publication continues with Issues and Challenges (Chapter 7), introduction to the Global TRIGA Research Reactor Network (Chapter 8) and concludes with Future Perspectives (Chapter 9). The publication summarizes in one compact form information on the past and present of TRIGA reactors and to give an outlook especially in view of potential issues to be solved by TRIGA operating organizations in the near future. As the staff, who participated in the construction and operated their TRIGA facilities throughout the decades have already retired in most of the institutions, there was an urgent need to gather their knowledge from already published sources or available as internal reports and documentation into a single publication. Further this publication is complemented with an attached CD-ROM, which includes a large number of individual papers, describing operational TRIGA facilities, reporting on their key research results, issues and challenges as well as future plans.

Speaker: Prof. Helmuth Boeck (Vienna University of Technology-Atominstitut)

Paper IAEA-CN-231-52.pdf

Slides 52_IAEA-CN-231-52_ppt_Rev.1.pdf

17:50 Challenges of Research Reactors Optimal utilization in Arab Countries

There are eight research reactors at present in the Arab world, one under construction, another one planned and two shutdown and decommissioned. The level of their operation and utilization differ from one country to another depending on the individual situation in a particular country. Some other Arab countries are constructing or planning to build new research reactors. These RRs are mostly used in: analysis of the structure of matter, radiation damage studies to develop better materials for nuclear and industrial applications, neutron activation analysis for accurate determination of elemental concentrations in material, production of isotopes that are used in biology, medicine, agriculture, industry, hydrology and research and training of scientists, engineers and technicians needed to support the nuclear power industry. The Arab Atomic Energy Agency (AAEA) is a regional specialized organization working within the framework of the League of Arab States to coordinate the scientific efforts of the Arab Countries in the field of peaceful uses of atomic energy. It contributes also to the transfer of the peaceful nuclear knowledge and technologies. One of the most important tasks of AAEA is to coordinate between Arab states to share their laboratory facilities and develop the human resources which have the capabilities of assimilating the nuclear knowledge and its application. The use of nuclear research reactors depends heavily on the availability of qualified scientists, engineers and technicians. Many Arab countries still have insufficient training capabilities in nuclear fields, and are experiencing problems with high staff turnover and shortage of specialized professionals in these areas. AAEA sponsored a coordinated research project put down by Arab experts according to the needs of sustainable development in Arab states and implemented within the human and technological resources available in the country and sharing of laboratory and technological capabilities with other AAEA member states. The project is accompanied by continuous cooperation between researchers and by human resources development and expert missions for the participating researchers and technicians in order to improve their skills and performances. The ultimate objective of the coordinated research project is to define and develop the preliminary steps and methods necessary to help in establishing a sound research and utilization program of available RRs in the Arab region. Many activities have been undertaken by AAEA related to the utilization of RRs such as; training courses, on-the-job training, training schools, scientific visits, scientific and experts meeting. Those activities cover a wide range of subjects related to RRs. Following are some of the training subjects undertaken regularly by AAEA: - Research reactors: Design, operation and applications. - Neutron Activation Analysis using RRs. - Reactor safety and security systems. - Radiation protection, regulations and legislations. - Emergency plans, waste management, monitoring and early warning. - Modelling of nuclear accidents and their effects on the environment and public health. - Workshops and fora about the applications of RRs. The research reactor is a very versatile tool, that when used effectively, can contribute to a country’s technological and scientific development. As most of the research reactor facilities are not being fully utilized, therefore AAEA regards that its technical cooperation project between Arab countries in the field of RRs utilization is of most interest on long-term sustainability of RRs utilization programmes. Therefore, countries which do not have a RR can benefit a great deal from these AAEA activities and enjoy the availability of facilities they do not have. Below we summarized the characteristics of the research reactors in Arab countries Country Facility Name Thermal Power (kW) Type Status Criticality Date Algeria ES-SALAM 15,000.00 HEAVY WATER OPER 1992/02/17 Algeria NUR 1,000.00 POOL OPER 1989/03/24 Egypt ETRR-1 2,000.00 TANK WWR OPER 1961/02/08 Egypt ETRR-2 22,000.00 POOL OPER 1997/11/27 Iraq IRT-5000 5,000.00 POOL, IRT SHUT 1967/01/01 Iraq TAMMUZ-2 500.00 POOL SHUT 1987/03/01 Jordan JRTR 5,000.00 TANK IN POOL UC Jordan JSA - Jordan Subcritical Assembly 0.00 SUBCRIT OPER Libya IRT-1 10,000.00 POOL, IRT OPER 1981/08/28 Libya Tajura Critical Stand 0.00 CRIT OPER 1981/08/28 Morocco MA-R1 2,000.00 TRIGA MARK II OPER 2007/05/02 Syrian Arab Republic SRR-1 30.00 MNSR OPER 1996/03/04 Saudi Arabia RR-1 POOL 30.0000 PLANNED Table 1: Status of Research Reactors in Arab Countries, including critical and sub-critical facilities.

Speaker: Dr Abdelmajid MAHJOUB (Director General (AAEA))

Paper ID_93.doc

18:10 → 18:30 Group Photo and Reception

Tuesday, 17 November

09:00 → 10:40 Safety of Research Reactors

Convener: Prof. Jose Lolich (Instituto Balseiro (Argentina))

09:00 IAEA SUB-PROGRAMME ON SAFETY ENHANCEMENT OF RESEARCH REACTORS

For more than sixty years research reactors has been a corner stone in the development and application of nuclear science and technology and in education and training of nuclear scientists and engineers. The benefits of research reactors also extended to many other fields including medical and industrial applications. Research reactors also can play an important role for building the technical and safety infrastructure in countries embarking on nuclear power programmes. Continued safe operation of research reactor facilities is essential for the benefits of the whole nuclear community. Through its programmes and activities, the IAEA continues to provide a main contribution to enhancing the safety of research reactors worldwide. The programme and associated activities are continuously adapted to address the needs of Member States, issues and trends, and challenges facing the research reactor community. The activities support Member States to improve their regulatory effectiveness and to enhance the management of safety of their facilities through the application of the Code of Conduct on the Safety of Research Reactors and the IAEA safety standards. The development and promotion of IAEA safety standards is a key activity within the sub-programme. At present, the corpus of safety standards reached maturity. Eleven Safety Guides have been published covering all areas important to safety, and significant progress has been achieved in the revision of the safety requirements documents to incorporate the feedback from the accident at the Fukushima-Daiichi nuclear power plants. Several Safety reports and technical documents were also published to provide additional guidance. These safety standards documents form the basis of the IAEA safety review services for research reactors, including the Integrated Safety Assessment (INSARR) missions. For Member States embarking on their first research reactor or a new reactor project, the IAEA provides a range of services to support Member States to establish the necessary infrastructure, including publication of guidelines, conduct of expert and advisory services and organization of training activities. The IAEA also continued to provide support to enhance the safety of research reactors under Project and Supply Agreements. Additionally, IAEA continued to provide support to research reactor organizations in a wide range of areas including regulatory supervision, safety management, safety analysis, ageing management, operational radiation protection, safety of experiments, emergency planning, decommissioning plans, and managing the interface between safety and security. Safety is further enhanced by the international exchange of information and sharing of operational experience feedback. The IAEA operates a web-based incident reporting system for research reactors (IRSRR) that facilitates the collection of information, analysis of data and dissemination of lessons learned to enhance safety. To further promote networking and sharing of experience, the IAEA supports regional and international cooperation to enhance safety, including facilitating establishment of regional safety advisory committees and assisting them to function effectively. Regional committees are now functioning ion Africa, Asia, and Europe. The implementation of the abovementioned programme and activities resulted in significant progress in enhancing the safety of research reactors worldwide, as reported by the Member States self-assessments in application of the Code of Conduct on the Safety of Research Reactors. However, further improvements are needed in some areas and efforts are still needed to address emerging challenges. These areas management of ageing research reactors, safety of experiments, periodic safety reviews, decommissioning planning, and, in view of the feedback from the accident at the Fukushima-Daiichi nuclear power plants, with respect to regulatory effectiveness, consideration of human factors, protection against external hazards, and emergency preparedness in particular for reactors with potential off-site consequences. The paper presents and discusses the elements mentioned above with a summary of the IAEA activities, recent progress and achievements for improving research reactor safety worldwide, and outlines of the strategy for implementing further improvements.

Speaker: Mr Amgad Shokr (Reserach Reactor Safety Section, IAEA)

Slides C01_Shokr.pptx

09:30 Highlights of Safety Enhancements of Research reactors Based on Safety Reassessments following the Fukushima-Daiichi Accident

Following the Fukushima-Daiichi NPP accident which occurred on 11 March 2011, specific safety reassessments (called Complementary Safety Assessments or Stress Tests) were initiated in many countries for research reactors on requests of the regulatory bodies or decisions of the operating organizations. Their main objective was to take into account the feedback and preliminary lessons learned from this accident, which cover technical and organizational aspects, including in particular the design of the facilities against extreme hazards associated with the site, the emergency preparedness and the regulatory oversight. The focus of the specific safety reassessments was mainly to evaluate the robustness of the research reactor facilities against extreme but credible hazards which are more severe than those adopted for the design, and to address the defense in depth, the performance of the basic safety functions as well as the continuity of the facility monitoring function in such conditions. The methodology established by the IAEA for performing post Fukushima safety reassessments of research reactors was aimed at ensuring consistency and avoiding non justified discrepancies in the approaches adopted in different Member States. This methodology is based on a deterministic approach and consideration of combination of events and possible interactions with other facilities at the site. The paper will present, on the basis of available information at the international level, a general overview on the safety reassessments performed for research reactors with a main focus on safety enhancements, including: • Modifications of the facilities to enhance their robustness against extreme external hazards, such as the reinforcement of reactor buildings against earthquakes, implementation of additional monitoring instrumentation including seismic detectors with associated action to shut down the reactor, implementation of an emergency control room from which it will be possible to shut down the reactor, to monitor important safety parameters and to control the operation of important safety systems in case of non-availability or non-accessibility of the main control room, provision of portable electrical power supply systems to operate items important to safety in case of a total blackout, implementation or improvement of emergency cooling systems to prevent potential fuel damage in case of a LOCA; • Consideration of Design Extension Conditions in the safety analyses and improvement of the consistency of analyses related to site specific extreme hazards; • Emergency plan and associated equipment and procedures for mitigation of severe accident consequences. The paper will present the above elements in detailed manner and include considerations on the use of a graded approach in performing the post Fukushima safety reassessments and on the follow up of implementation of the resulting safety enhancements.

Speaker: Dr HASSAN ABOU YEHIA (Institut de Radioprotection et de Sûreté Nucléaire, Nuclear Safety Division)

Paper Invited_Paper-H._ABOU_YEHIA.pdf

10:00 STATUS OF JRR-3 AFTER GREAT EAST JAPAN EARTHQUAKE

JRR-3 at Tokai site of JAEA was in its regular maintenance period, when the Great East Japan Earthquake took place on 11th March 2011. The reactor building with the solid foundations and the equipment important to safety survived the earthquake without serious damage, and no radioactive leakage has been occurred. Recovery work, check and test of the integrity for all components have been carried out. In response to the accident at Fukushima Daiichi NPS, the new safety standards for research and test reactor facilities came into force on December 18, 2013. The evaluation of natural disasters and prevention of spread of accidents beyond design basis mainly were enhanced in the standards. We have completed the necessary checks and assessments, and submitted an application for reviewing if JRR-3 complies with the new standards to the Nuclear Regulation Authority on September 26, 2014.

Speaker: Mr Kato Tomoaki

Paper Full_Paper_(Kato_JAEA_Japan).pdf

Slides C03_Kato.pptx

10:40 → 11:10 Break: Poster Session and Coffee Break

11:10 → 12:50 Safety of Research Reactors

Convener: Dr Hassan ABOU YEHIA (Institut de Radioprotection et de Sûreté Nucléaire (IRSN) - FRANCE)

11:10 Safety Assessment and up-grades executed and proposed at Dhruva Reactor Following the Accident at Fukushima Daiichi Nuclear Power Plant

Dhruva is a 100 MW research reactor fuelled with natural uranium and cooled and moderated with heavy water. It is located in Bhabha Atomic Research Centre, Mumbai, it was commissioned in 1985. Following the Fukushima Daiichi nuclear accident a complete review of existing system configuration was undertaken to ascertain reactor’s capabilities to withstand extreme natural event, like earthquakes, tsunamis, storm surges etc. Various safety up-grades have been undertaken or are proposed to be executed in order to address issues which have surfaced. Several measures for improving safety margins against BDBEs, availability of the reactor and its effective utilization have been taken. After shutting down of the ‘Cirus” reactor, it is the lone reactor which is meeting the isotope requirements of the nation. During the recent years, reactor has attained its highest availability and capacity factor, reactor utilization has also been enhanced, to meet the national requirement of radioisotopes in quality and quantity. The paper will highlight the changes implemented and proposed in Dhruva for meeting these requirements.

Speaker: Mr Chandrashekahr Karhadkar (Bhabha Atomic Research Centre)

Paper IAEA-CN-231-7_Paper.pdf

Slides C04_Karhadkar.pptx

11:30 French Post-Fukushima Complementary Assessments – General Approach and Resulting Safety Improvements for the High Flux Reactor located in Grenoble

Following the accident that occurred on the Fukushima Daiichi nuclear power plant on the 11th March 2011, the French Prime minister asked the national nuclear safety authority (ASN) to engage a targeted reassessment of the safety of every basic nuclear facility with the aim of evaluating their capacity to withstand extreme situations beyond design basis assumptions. These specific reassessments, called Complementary Safety Assessments (CSAs), were mainly carried out by operators in 2011 on the basis of the specifications for the stress tests requested by the European Council. In France these evaluations included all nuclear power plants in operation but also nuclear cycle facilities and research reactors. The CSAs have been carried out with the purpose of analyzing the robustness of the nuclear facilities regarding : • natural hazards more severe that ones retained for its design; • specific situations such as long-term loss of cooling and loss of electrical power supplies. The paper will present the analysis performed by french operators in the framework of CSAs and the opinion of the Institute of Radioprotection and Nuclear Safety (IRSN) which has been largely involved in the reviewing of the CSAs carried out by licensees. In particular the paper will present the opinion of the IRSN on the assessment of the robustness of facilities by operators. Then, the paper will introduce the concept of “hardened safety core” firstly defined by IRSN on the basis of the conclusions of its CSAs critical review. The “hardened safety core” is a set of structures and equipment identified to withstand natural extreme hazards as earthquake and flooding, completed by crisis management measures. The identification of provisions to be included in “hardened safety core” must be done with respect to the “defence in-depth” principle considering the prevention of the occurrence of accident situations, whilst assuming they may still occur, thus envisaging the implementation of appropriate systems to manage them including provisions for crisis management actions. For research reactors, the definition of a “hardened safety core” must consider the potential of danger of each facility (graduated approach). Finally the paper will present a concrete application of the a “hardened safety core” based on the example of the High Flux Reactor (RHF) research reactor, operated by the Laue-Langevin Institute, at Grenoble (France). The RHF is an interesting example firstly because the reactor is located in a strong natural hazards area (high earthquake and flooding risks) and secondly because of the relative proximity of the ILL site from urbanized zones. The paper will focus on the modifications suggested by the operator and the conclusions of the review of these propositions done by IRSN between 2012 and 2014. The improvements already achieved in the facility and those that are still in progress will be presented, showing how these improvements answer to the objective of controlling the main vital safety functions (cooling of the reactor, confinement of radioactive materials, crisis management) in exceptional - but nonetheless conceivable - situations.

Speaker: Mr Emmanuel GROLLEAU (Institute of Radioprotection and Nuclear Safety (IRSN))

Paper ID_34.pdf

Slides AIEA-CN-231-34.PPT

11:50 Implementation of Regulatory guidelines based on IAEA SRS-80 to reassess and enhance the safety of SNRC IRR-1, 5MW Pool Type MTR

N. Hazensprung (1), T. Makmal (1) and Y. Barnea (2) (1) IRR-1, SNRC, Yavne, ISRAEL, 88000. (2) NLSO/IAEA, Tel-Aviv, ISRAEL, 61070. Following the accident in the Fukushima-Daiichi, Japan NPPs (March 2011), the previous ICRR in Rabat, Morocco (November 2011), recommended that the operating organizations will take a proactive approach to examining their design basis and safety analysis to evaluate what, if any, changes and improvements should be made to withstand multiple severe external events as appropriate for their site and facility characteristics. The paper presents an overview of the activities done on this topic by the ISRAEL Atomic Energy Commission (IAEC) Nuclear Licensing and Safety Office (NLSO), and the operator of Soreq-Nuclear Research Center (SNRC) IRR-1, 5MW, Pool-Type MTR. Although some of the activities started prior and aftermath of the accident, most of the regulatory instruction and operator’s implementations were following the draft of IAEA SRS-80 technical guide, formally published on March 2014. The activities started with a specific regulatory decree (requirements), presented on March 2013 to the operator. The requirements were transformed to an action plan and included: a) reassessment of the calculated PIE’s in the SAR, and including combined external hazards, specifically with fire; b) upgrading the reactor safety systems and safety related systems upon the safety review; c) re-evaluation of site seismic database and analysis; d) auditing the analyses on the robustness of preselected SSCs, mainly those related to the nominal and emergency power supply systems and cooling system; e) review of the completeness and updating the operational procedures; f) review of the emergency procedures and drills. The main part of the document describes the activities accomplished during a less than three years program, to be completed through 2015. A notable safety upgrade to be mentioned is the connection of two different ground accelerometers to the reactor protection system enabling automatically shutdown the reactor during an earthquake.

Speaker: Dr Yacov Barnea (NLSO - IAEC)

Paper 152_C06_Barnea.docx

Slides C06_Barnea.pptx

12:10 SAFARI 1 SAFETY REASSESSMENT OUTCOME AND MODIFICATIONS IN LIGHT OF THE FUKUSHIMA DAICHI ACCIDENT

Following the Fukushima nuclear accident, a directive from South Africa’s National Nuclear Regulator was received which required a safety reassessment of the SAFARI 1 research reactor. The safety reassessment consisted of: • Evaluation of the response of the SAFARI 1 Research Reactor when facing a set of extreme external events (EEE) and • Verification of the preventive and mitigation measures chosen following a defence-in-depth (DiD) logic: initiating events, consequential loss of safety functions, severe accident management. The safety reassessment process was performed in various steps. Site-specific natural external events were firstly identified. The full lists of EEEs identified that may have an impact on SAFARI 1 include earthquakes, external flooding, tornadoes and tornado missiles, high winds, sandstorms, storms and lightning, hurricanes and tropical cyclones, bush fires, explosions, toxic spills, accidents on transport routes, effects from adjacent facilities, biological hazards, and power or voltage surges. This step was followed by the development of event trees which depict the progressive evolution of the EEE into plant damage states which could potentially lead to public exposure. These evaluations were carried out in accordance with the philosophy of DiD as proposed in the ENSREG stress test specification. This paper will present the feasibility phase outcome, results of the safety reassessment, as well as some of the resulting modifications and the future plans to conclude the post-Fukushima activities.

Speaker: Mr SM MALAKA (NECSA)

Paper Paper_SAFARI_1_SAFETY_REASSESSMENT_OUTCOME_AND_MODIFICATIONS_IN_LIGHT_OF_THE_FUKUSHIMA_DAICHI_ACCIDENT__SM_MALAKA.pdf

12:30 Thermal Hydraulic and Safety Analysis for the Moroccan TRIGA MARK II Research Reactor

In this study we have calculated some important thermal–hydraulic parameters of the Moroccan 2-MW TRIGA MARK operating under normal condition. We have also presented the analysis of some abnormal situation; the small break loss-of-coolant transients in this pool-type research reactor, with SCRAM disabled, is simulated based on the PARET model previously established and validated for our TRIGA reactor. The study involves the determination of the Departure from nucleate boiling ratio (DNBR), the temperature profile and heat flux across the hottest channel. The results indicate that the peak clad temperatures remain well bellow the clad melting temperature for small break LOCA accident during long time of these transients and the reactor will have sufficient safety margins against this abnormal situation.

Speaker: Dr Yassine Boulaich (CNESTEN)

12:50 → 13:50 Lunch Break

13:50 → 15:30 Operation and Maintenance

Convener: Mr Benji Steynberg (SAFARI-1 Necsa)

13:50 IAEA Activities in Support of Research Reactor Operation and Maintenance

Currently there are 247 operating research reactors globally, according to the International Atomic Energy Agency (IAEA) Research Reactor Database (RRDB). These reactors have a well-documented history of contributing to peaceful nuclear research and technology development, and have helped in the education and training of generations of scientists, reactor operators, and engineers. They are also used for basic research, radioisotope production, neutron radiography, neutron beam research, material characterization and testing, and other applications. The IAEA Research Reactor Section (RRS) has several initiatives to assist Member States (MS) individually, regionally, and through international coalitions with operational and maintenance (O&M) issues. Some of the key issues related to RR O&M are due to ageing facilities and equipment. IAEA is leading several efforts, through Coordinated Research Projects and development of publications, to enable MS to share knowledge about material ageing and available equipment and facility upgrades to sustain RR operability. Additional O&M issues being addressed by MS are fuel optimization, equipment modernization, modifications required due to security and safety requirement changes, and modifications aimed at increasing facility reliability. The RRS offers MS Operations and Maintenance Assessment of Research Reactors (OMARR), a peer-to-peer review to assist with improvement of operational and maintenance practices. Thus far, two facilities have used this opportunity, and another is planned for 2016. IAEA is establishing a specialized activity for conducting non-destructive examinations and in-service inspections at research reactors. Additionally, RRS is currently participating in several projects through the Technical Cooperation organization to assist individual and regional MS on specific projects. These relate to instrumentation upgrades, fuel upgrades, safety infrastructure support, and decommissioning planning. This presentation will provide information on the current and planned IAEA RRS activities to support MS with RR O&M.

Speaker: Frances Marshall (International Atomic Energy Agency)

14:20 RELIABILITY IMPROVEMENT AT OPAL

OPAL is a 20MW(t) reactor facility operated by the Australian Nuclear Science and Technology Organisation (ANSTO) in Sydney, Australia. Commissioned in 2006, OPAL is a multipurpose reactor used to perform a range of commercial and scientific irradiations supporting ANSTO’s radiopharmaceutical production and industrial irradiation businesses, as well as providing thermal and cold neutron beams to support ANSTO’s neutron science programs. Increasing demands being placed on the reactor by stakeholders from both commercial and research sectors has led to a focus on achieving sustainable improvements in reactor performance and reliability while continuing to manage safety, risk and regulatory compliance. These demands are expected to further increase in 2016/17 with the commissioning of the ANSTO Australian Nuclear Medicine (ANM) Facility which is set to triple ANSTO’s current production of Mo-99. This paper will outline some of the strategies and processes adopted by OPAL to manage and improve plant performance and reliability. Examples of specific topics include asset management framework, maintenance and capital investment strategy optimisation, proactive maintenance programs such as condition monitoring, lubrication management, precision maintenance and operator driven reliability, reliability engineering, ageing management, shutdown management, risk management, integrated logistics support and maintenance information systems.

Speaker: Mr Andrew Frikken (ANSTO)

14:50 Management of Ageing and Modifications for research reactors

Egypt has two research reactors, the first reactor was designed by the Former Soviet Union, it has a power of 2MW, tank type, light water moderated and reflected and the fuel type is EK-10 which consists of Uo2 fuel rods with 10% enrichment. The reactor was commissioned on 1960 and was working until 2010. After that it has been in an extended shutdown state due to ageing of mechanical systems and the termination of the fresh fuel in the international market and for safety considerations. Based on the in-service inspection which was implemented for the 1st reactor in 1999, the following systems were modernized: 1. Computerized Safety Logic System. 2. Process parameters measurements 3. Neutron parameters measurements 4. Control circuits for the movement of control rods. 5. Control circuits for opening and closing the horizontal channels 6. Power supply systems. (transformer and circuits) 7. Radiation protection system. Some of these modifications were implemented through technical assistance with IAEA and others by internal bids from Hungary or France or Egyptian companies. A brief description and the knowledge obtained from each modernized activity will be presented, showing the steps which were implemented to achieve the high performance and availability of the first reactor during its operation period. The second reactor is a Material Testing Reactor (MTR) type using plate type fuel elements with 19.75% enrichment and power equal to 22 MW. It went critical in 1997, and used mainly for production of Mo-99 from Low Enriched uranium (LEU) targets. Ageing is defined as a process in which characteristics of components, systems and structures gradually change with time or use. The service conditions can be a major factor for ageing. In order to achieve the overall safety objective for a research reactor the defence in depth concept and multi barriers should be preserved. The ageing process can accelerate the failure probability of a barrier component and ultimately to the failure of the barrier. The ETRR-2 reactor has a good ageing management program for all the mechanical, electrical, Instrumentation and control systems. This program includes; periodic maintenance activities, periodic calibration and testing program, in-service inspection. The program should meet the safety requirements from: • General safety objectives for research reactor to protect individuals, society and the environment. • Safety analysis report (SAR). • Operational limits and conditions (OLCs). • Regulatory body. • Reactor management. • Manufactures recommendations Also, there is a program to avoid the corrosion in the ageing cooling system by using special chemicals in order to preserve the water quality in the permissible limits. The outlines of ageing management program to achieve the high performance, reliability and availability for the second Egyptian research reactor consists of: • Preventive maintenance, Predictive maintenance, Corrective maintenance, • Repairing and replacing of components, • Maintenance procedure and format and Work permit, • Service conditions and its effects on ageing, • Ageing mechanisms which lead to failure of components or systems, • Maintenance report: o weekly maintenance performance report o performance analysis report, • Surveillance and review program, • Examples of practical safety related events during the last period of operation due to ageing problems or deficiency in the management of ageing program. • Safety Committee Review and assessment of the maintenance program before sending it to the Regulatory Body for approval.

Speaker: Prof. Mohamed Shaat (Professor)

Paper Dr_Shaat-_Management_of_Ageing_and_Modifications_for_____Research_Reactors_(final)(1).pdf

Slides IAEA-CN-231-12,_Management_of_Ageing_and_Modifications_for__RRs,_PP._Presentation.pptx

15:10 FRM II: EXTENSION OF CORE COMPONENT LIFETIME BY APPLICATION OF FRACTURE MECHANICS

The FRM II is Germany’s most modern research reactor. It became critical for the first time on March 2nd, 2004. Since routine operation started in May 2005, it has just completed its first ten years of operation. The FRM II is a multi-purpose reactor and operates at nominal power of 20 MW. It is light water cooled and heavy water moderated. It is used predominantly for neutron scattering experiments in a wide range of applications from fundamental physics to material and life sciences but also runs a prompt gamma neutron activation analysis facility, a tomography beam line, a positron source and a medical irradiation facility. Important fields of activity include also isotope production and Silicon doping. Most of the core components of the FRM II are made from Aluminium (EN AW-5754, AlMg3). This material undergoes embrittlement under neutron irradiation, mainly by Silicon formation in the Aluminium by capture of thermal neutrons. Since only few data on highly irradiated Aluminium are available worldwide the FRM II runs its own irradiation programme. This is also a requirement of the licensing procedure. Based on the concept of tensile strength samples of the same Aluminium are irradiated during the operation. After accumulation of certain neutron fluences the samples are removed and the remaining tensile stress as well as fracture mechanics parameters are determined. The concept of using tensile stress is the only one accepted for the FRM II so far. Although it is a well-established procedure, it is in the case of the FRM II core components very conservative and not necessarily in line with the latest knowledge in science and technology. Consequently, it may lead to the requirement of exchange of some components before they reach their true end of lifetime. This, in turn, unnecessarily reduces the availability of the FRM II due to additional maintenance breaks and generates avoidable radiation dose for the personnel. Therefore together with external fracture mechanics experts, the expert organization and the licensing authorities a different concept has been developed: The proof of sufficient stability shall now be performed using fracture mechanics. To this end, detailed calculations on the mechanical stress on the affected components have been carried out. These components are mainly the central channel that houses the fuel element, the moderator tank itself, the beam tubes as far as they are located in high neutron flux areas and other components made form Aluminium that reach into the moderator tank and into areas with high neutron flux. Maximum undetected fracture size and fracture growth have been assumed in application of the relevant codes. Detailed neutron fluence calculations have been carried out and parameters on the material properties after irradiation in the light of fracture mechanics were deduced from the samples irradiated at the FRM II. Finally all these results were put together leading to a detailed picture of required material properties – as necessary for safe operation – and existing material properties – after a certain irradiation was accumulated – which lead to a prediction of the remaining lifetime of the core components made from Aluminium. Although the final result has not been approved in every detail by the licensing authority already now a premature exchange of some components could be avoided. The complete new concept for the lifetime of core components made from Aluminium, based on analyses of irradiated samples, fracture mechanics and detailed fluence calculations will be available soon.

Speaker: Mr Axel Pichlmaier (TU-München, FRM II)

Paper Pichlmaier.pdf

15:30 → 16:00 Break: Poster Session and Coffee Break

16:00 → 17:50 Operation and Maintenance

Convener: Mr Axel Pichlmaier (TU-München, FRM II)

16:00 The development of high density, low enriched fuel for the conversion of research reactors

The development high density LEU fuels was initiated by the U.S. Reduced Enrichment for Research and Test Reactors (RERTR) program, established in 1978 to address a concern about the proliferation of HEU in civil commerce. Its goal was to enable research and test reactors to convert to, or back to, LEU by developing higher-density fuels that could accommodate the required increase in 238U and that could be used without significant performance loss or cost increase. Initially the program concentrated on the densest uranium silicide phase (U3Si) dispersed in aluminium as the prime candidate for development. The next-lower-density silicide phase, U3Si2, was also tested. Rapidly international cooperation started up for development of the silicide dispersion fuels. However the main silicide fuels being tested for use in plate-type fuel elements, U3Si-Al and U3SiAl-Al, had shown fission gas bubble growth in the fuel particles by at ~ 85% burn up in LEU. Fortunately, U3Si2-Al fuel was found stable at a burn up of at least 90%, so further development was concentrated on U3Si2-Al fuel. The development of this fuel evolved without major drawbacks, and in 1988, the NRC gave a generic approval for use of U3Si2-Al dispersion fuel at 4.8 gU/cm3 in its licensed research and test reactors. This fuel gained worldwide acceptance and many research and test reactors, with powers up to 50 MW, have been converted since 1988 using U3Si2-Al fuel. In Canada AECL developed and qualified U3SiAl-Al and U3Si-Al dispersions for its reactors, which use pin-type fuel elements. Later, AECL also qualified U3Si2-Al fuel in MAPLE-type pins. The one negative aspect of the silicide fuels was the formation of silica gel during dissolution of spent fuel during reprocessing. - The problematic of the closure of the LEU fuel cycle will also be addressed in the paper. High-power-density research and test reactors could not be converted with U3Si2 as they require higher-density fuels. A fuel loading of 6.5-8.5 gU/cc is required for most of these reactors; some of them require even higher densities only achievable with alternative fuel designs, such as the monolithic UMo. So the fuel developers community started the development of very-high-density fuels. High-uranium-content alloys, such as UMo and UNbZr with uranium contents of 90 w% or higher were considered. Guided by past experience with fast reactor fuels, UMo was adopted as the primary candidate for a very-high-density fuel by the U.S. RERTR program. The initial irradiation tests of UxMo dispersions showed the fuel to perform satisfactorily with x ≥ 6. There was however considerably more interaction between the fuel and the aluminium matrix than there had been for the silicide fuel, but it was thought that this problem could be overcome. Sometime later the U7Mo dispersion fuel system qualification experienced a number of unexpected setbacks, mainly related to the formation of an interaction layer around the fuel grains. The particular power regime at which the high density U7Mo fuel needs to operate, present a particular challenge compared to the U3Si2 qualification process. Several remedies for this failure mechanisms were proposed and tested. Silicon addition to the matrix or coating of the fuel grains (E-FUTURE and SELENIUM) have clearly improved the behaviour up to 60% burn up. However, beyond these fission densities, fuel plates are still found to show rapid swelling. The tests with coated UMo fuel also show an accelerating swelling in function of burn up at these elevated fission densities, but were not followed by pillowing. The current working hypothesis is that we are facing intrinsic properties of the atomised U7Mo fuel, where the onset and progress of recrystallization of the fuel may eventually cause swelling rates to exceed the mechanical capabilities of the irradiated dispersion UMo meat. The recrystallization threshold and evolution can be influenced by changing the UMo microstructure to eliminate Mo inhomogeneity and enlarge grains, which may hold promise to reduce the swelling rate at high burn up to within acceptable limits for the mechanically weakened matrix. Presently both dispersion and monolithic fuels are being pursued in parallel. Qualification of UMo dispersion fuel is being pursued by the European HERACLES program. The U.S. program has a collaborative relationship with the HERACLES program. Two irradiation campaigns are presently foreseen to test and optimize the various remedies and deepen the basic understanding of the failure mechanisms. The fabrication process may have to integrate additional steps: atomization, coating, heat treatment. Qualification of UMo monolithic fuel is strictly within the U.S. program at the present time, aimed at the conversion of five U.S. research and test reactors. The efforts are mainly concentrated on the industrialization of the fabrication process.

Speaker: Dr Edgar Koonen (SCK-CEN)

Paper ID_104.pdf

16:30 Continuing Progress toward Research and Test Reactors Conversion to Low Enriched Uranium Fuel

Nuclear research and test reactors have been built and operated across the globe to serve many missions in science, engineering, and medicine. Over time, many of the reactors have been operated with High Enriched Uranium (HEU, with U-235 enrichment ≥ 20%) fuel. In response to concerns about the potential misuse of HEU, the U.S. Department of Energy (DOE) initiated a program – the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. In 2004, the reactor conversion program became the driving pillar of the Global Threat Reduction Initiative (GTRI), a program established by the U.S. DOE’s National Nuclear Security Administration (NNSA) to minimize, and to the extent possible eliminate, the use of HEU in civilian commerce. As an integral part of the GTRI, the Conversion Program accelerated the schedules and plans for conversion of additional research reactors operating with HEU. In 2015, the NNSA weapons-usable material efforts were realigned into the Office of Material Management & Minimization to further integrate permanent threat reduction. A summary of the technical aspects of the conversion program was presented at the previous IAEA International Meetings on Research Reactors in 2007, and 2011. This paper provides an update on the progress made since 2011 and describes current technical challenges that the program faces. Since the inception of the Conversion Program, 66 research reactors have been converted to LEU fuel and 25 have shutdown prior to conversion. Furthermore, an HEU-target fission-product Mo-99 producer has converted to the use of LEU targets. The total count of 92 facilities no longer using HEU is an increase of 13 since the Rabat conference in 2011 (i.e., 16% more than the 79 reported in 2011). A key element for the success of the program is the establishment of international collaborations, especially with the IAEA. The major technical activities of the Reactor Conversion Program include: (1) the development of advanced LEU fuels and (2) conversion analysis and conversion support. The most challenging factor in enabling the conversion of a research reactor lies in the availability of an LEU fuel with density high enough to compensate for the reduction in the content of U-235 and the associated absorption of neutrons in the increased U-238 in the LEU material. The conversion of all U.S. civilian research reactors that were able to convert with existing LEU fuel was completed in 2009. The remaining research and test reactors in the U.S. are all high flux reactors (U.S. High Performance Research Reactors - USHPRR) and require higher density fuels not currently available. Significant efforts have been ongoing for the development of a monolithic form of Uranium-Molybdenum (UMo) alloy fuel that provides densities of U high enough to enable the conversion of the U.S. high flux reactors. In addition, the Conversion Program has long been engaged in cooperation with Russian organizations to develop high density dispersion UMo fuel for use in Russian designed reactors. Finally, the Conversion Program has a strong partnership with the European effort LEONIDAS that has grown into the HERACLES Group for the development of high density dispersion UMo fuel for the conversion of the European High Flux research reactors that use HEU fuel. The analysis and support activities provide the required analytical and design evaluations to support the conversion of research reactors involved in the program. Since the inception of the program, analysis methods and codes have been developed specifically for the analysis of research reactors. The methods and codes are currently evolving to incorporate the latest tools and data, with an increased emphasis on validation with experimental data. Conversion analysis in general includes three major tasks: ● Feasibility studies to determine suitable LEU fuel assembly designs which minimize performance impacts for each reactor. ● Operational and safety analysis, necessary to demonstrate that the transition from HEU to LEU fuel can be done safely and without interrupting normal operations ● Resolution of regulatory issues to obtain regulatory approval for the conversion to LEU fuel. It must be demonstrated that all safety requirements are met. A project established through the IAEA has moved the previous cooperative project for determining the feasibility of converting Chinese-designed MNSR reactors to LEU fuel, into a project for the actual fabrication and testing of fuel and conversion of the reactors. The first LEU conversion core for an MNSR was tested successfully in the new Zero Power Test Facility (ZPTF) and will be used to convert the prototype MNSR-IAE in Beijing. The paper will provide a more detailed overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives.

Speaker: Dr John Stevens (Argonne National Laboratory)

16:50 WWR-K Reactor Upgrades Related to HEU/LEU Conversion

WWR-K is a light water 6 MW research reactor operated by the Institute of Nuclear Physics in Almaty, Kazakhstan. Activities on conversion of the reactor from HEU to LEU fuel are in progress. LEU fuel assembly for WWR-K has been designed and tested. Necessary modifications of some reactor systems are carried out in frame of the conversion programme. ­ replacement of instrumentation and control (I&C) system electronics; ­ replacement of control rods, control rod drives, channels and support structure for mounting the channels; ­ replacement of emergency cooling system with an uninterruptible power supply and the emergency sprinklers for the reactor core; ­ replacement of radiation monitoring system; ­ installation of gas and aerosol emissions monitoring system; ­ installation and testing of additional new cooling towers; Also, a detail inspection of the reactor tank and primary circuit piping has to be done. Startup of the WWR-K reactor with LEU fuel is expected in the end of 2015. This work is supported by the US Department of Energy and Ministry of Energy of Kazakhstan.

Speaker: Mr Petr Chakrov (Committee of atomic and energy supervise and control)

17:10 The study and analyze the operation of the Bandung TRIGA research reactor using plate type fuel elements.

In order to support the national development program, the National Nuclear Energy Agency of Indonesia has among other duties is to conduct activities of research, development, radioisotope production and utilization of nuclear energy. To carry out these duties, the Bandung TRIGA reactor operation should be considered. Until recently, the Bandung TRIGA reactor still has strategic and economic value, particularly to back-up radioisotope production of the Serpong reactor and to conduct research in the field of reactor physics, reactor thermal-hydraulic, reactor instrumentation, research on neutron activation analysis and other research. Nevertheless, the operation of the Bandung TRIGA reactor is dependent on the supply of standard TRIGA fuel, but the company for fuel suppliers for standard TRIGA reactor does not exist anymore. While on the other hand BATAN Serpong has been able to make its own fuel plate type and it has been used for Serpong reactor. BATAN Serpong has been able to produce plate-type fuel, and it has anability to perform analysis onthermal-hydraulicand neutronic for this type of fuel. It also has experience in the design of control systems and computer-based instrumentation of research reactor; it can be beneficial for BATAN to keep the Bandung TRIGA reactor remain in operation in Indonesia by using fuel plate type.Therefore, BATAN researchers particularly at the Center of Applied Nuclear Science and Technology (CANST) Bandung and some associated centers have plan to make of the Bandung TRIGA reactor conversion by changing the type of fuel rods into fuel-type plate, by using of an artificial BATAN fuel element itself, so the Bandung TRIGA reactor fuel supply does not depend on the standard TRIGA fuel. This activity will also be able to show that the BATAN been able to operate the reactor independently such as Bandung TRIGA reactor. Strategies to achieve results to obtain the type plate Bandung TRIGA reactor fuel is to do some design and analysis, such as reactor core design to get the optimal core. To determine the operating cycle of the TRIGA reactor using plate-type fuel, necessary burnup calculation of each cycle and how to perform core management optimally. Currently being conducted the design of the primary and secondary cooling system of the reactor, the design of the delay tank system, and the design of the reactor instrumentation and control systems. For the cooling system, because the flow of coolant in the reactor core will use the forced convection flow, needs analysis and new design of the primary coolant system. This design is based on the data flow rate is required, so that the design of the reactor primary coolant system produced in accordance with the conditions of the planned reactor core. In the planning the Bandung TRIGA reactor will be operated at maximum power of 2000 kW. A delay tank will be designed to inhibit trips N-16 that will disintegrate before the cooling water back into the reactor tank and N-16 has decayed. Implementation of the program began in 2015 and it was estimated to obtain the final design in 2019. In addition, safety studies related to the project are being performed in order to obtain a new license for the entire facility, not only because of the modifications but also to generate updated regulatory documents, including the safety analysis report. At this time, because BATAN Bandung still has the standard TRIGA fuel, the Bandung TRIGA reactor operated using standard TRIGA fuel to the end of 2018. Because of its FFCR has burned nearly 50%, the reactor control rods will be replaced by using control rods without FFCR. Control rods with out fuel followers have been made in BATAN Serpong and now are in the testing phase. Keywords: Bandung TRIGA reactor, plate type fuel, FFCR

Speaker: Mr Ketut KAMAJAYA (National Nuclear Energy Agency (BATAN) of Indonesia)

Paper Thermal neutron flux 2.89 x 1013 cm-2.sec-1 on the center irradiation position (C-3) and 1.55 x 1013 cm-2.sec-1 on irradiation position (A-1, A-5, E-1, and E-5)

Slides IAEA-CN-231-55_Kamajaya.pptx

17:30 Calculations and Measurements for the Full-Core Conversion of the WWR-M Research Reactor in Ukraine

In accordance with the program of pilot usage of LEU fuel approved by the Ukrainian Regulatory Committee, most burned HEU fuel assemblies of the WWR-M research reactor were successively replaced by fresh LEU fuel. By using this way, neutronic performance of the reactor remained almost the same as with HEU fuel but such the conversion progressed very slowly. Thus, the new full-core conversion program with simultaneous replacement of all remaining HEU fuel by fresh LEU fuel was developed. The models applied for the conversion safety analysis were validated against measured data, which included critical experiment results for fresh fuel assemblies and measured neutronic distributions in real WWR-M reactor core. However, because of essential decrease of the number of fuel assemblies in the core (from 210 to 72), a lot of beryllium blocks had to be loaded into the reactor core. Most of these blocks were not used more than 40 years and information of their irradiation history was not available. Thus, excess reactivity for the new LEU core was difficult to calculate accurately because of unknown He-3 poisoning. To provide safety of the new core loading, conservative approach was used. Irradiated beryllium blocks with unknown He-3 poisoning were assumed to be fresh, and 15 aluminum blocks were loaded for the nonce instead of beryllium to decrease excess reactivity. Moreover, neutron flux was being monitored all the time during the core loading to estimate subcriticality and worth of the control and safety rods. When criticality was reached, excess reactivity and He-3 poisoning were estimated. Since He-3 poisoning was estimated to be high, 15 temporary aluminum blocks were replaced by beryllium ones. Then worth of the rods and excess reactivity were measured. By using comparative reactivity measurements, the beryllium blocks with the highest poisoning were detected and moved far away from the fuel to diminish their influence on the neutronics and thermal-hydraulics parameters of the core. Since such beryllium shuffling changed the worth of the rods and excess reactivity essentially, they were measured again. This measurement was in good agreement with calculation, so safety analysis was validated for the new LEU core. Drop of the reactor power and number of fuel assemblies in the new core resulted also in essential decrease of fast and intermediate neutron flux in beam tubes. To solve this problem, LEU core configuration was optimized. Moreover, dependence of the number of fuel assemblies in the core and maximum allowed power of the reactor on LEU fuel burnup was estimated. Using this dependence, it was estimated how transient core configuration should be changed with LEU fuel burnup to provide sufficient fast and intermediate neutron flux in beam tubes. Then, placement of fuel assemblies and beryllium blocks in the core was optimized for each core reload during transient period to satisfy all the safety requirements and provide high neutronic performance of the reactor.

Speaker: Dr Yuri Mahlers (Institute for Nuclear Research)

Paper Calculations and Measurements for the Full-Core Conversion of the WWR-M Research Reactor in Ukraine

17:50 → 18:00 Coffee Break

18:00 → 19:00 Side Event

Wednesday, 18 November

09:00 → 10:30 New Research Reactor Project

Convener: Mr Alexander Tuzov (Joint Stock Company "State Scientific Center - Research Institute of Atomic Reactors")

09:00 Feasibility Study for a New Research Reactor Project

The number of new comer countries interested in developing a program to introduce a first research reactor for them has grown significantly. In addition, the number of projects to introduce a replacement research reactor or to expand the national research reactor capacity has increased as well. Thus, a guideline to insight the overall process of a research reactor project has been asked and the IAEA published a guidebook in 2012, which is often called a milestone document [1]. This guide specifies three phases before operation and they are Pre-project phase, Project formulation phase and Implementation phase. For the phase 1, i.e., pre-project phase, a preliminary strategic plan should be prepared and this report should show that there are sufficient needs at national level that justify the research reactor project. Another output of phase 1 should be a feasibility study report which demonstrates that a nation or an organization is in a position to make a decision whether to proceed with the new research reactor project or not. This report will show all the obligations and commitment involved and should include a long term national strategy. The activities for a feasibility study or the contents of a feasibility study report may depend on countries, stakeholders or backgrounds of research reactor projects. However, basically, the feasibility study should include the analysis of cost, benefit and risk involved in the realization of the results of the strategic plan. In addition, the feasibility study should include a comprehensive assessment of all 19 national infrastructure issues described in the Milestone document. The benefit from a research reactor may be a direct benefit such as the revenue from the sales of products or services. For a research reactor, the most of the benefit may be indirect benefit such as the contribution to the basic research and the contribution to training and education. Also, there are many different ways to count the indirect benefit. The choice of analysis method will be a decision of a study team which should depend on the background of a research reactor project. As for the cost analysis, the project cost, the cost for operation and maintenance, the fuel cycle cost and the decommission cost should be included. The difficulties in the cost and benefit analysis are in that the study should predict the life time of a facility and in that a research reactor is a custom-design product which makes it very difficult to find good references for the cost estimation. The best way to overcome these will be to make many experts involved in the analysis and to have an enough time period for sharing idea and discussion. The risk analysis should involve the analysis of technical, social as well as financial risks. As for new comers, the establishment of the infrastructure at a time will be impossible and costly. Thus, a step-by-step policy may be taken and will be realistic. The feasibility study may be conducted by an operating organization or an independent organization depending on the decision or a rule of a country. An important aspect in this is that the technical capability as well as the independency of the study should be considered. What is more important than the analyses of cost, benefit and risks is believed to be the strategic decision or the intension for a project. If a research reactor project is strongly recommended and supported, many good ideas will be proposed during the feasibility study and will be accepted for a positive decision. This will be also a key factor for the success in the next phases until the operation phases. References [1] IAEA, NP-T-5.1, ”Specific Considerations and Milestones for a Research ReactorProject”,2012.

Speaker: Dr In-Cheol LIM (Korea Atomic Energy Research Institute)

Paper iclim_ICRR_2015_No53.docx

09:30 THE ROLE OF RRS IN THE DEVELOPMENT OF THE NATIONAL INFRASTRUCTURE FOR NUCLEAR SCIENCE AND TECHNOLOGY PROGRAMMES

The main purpose of this paper is to define the potential and limitations of a Research Reactor (RR) and related nuclear infrastructure in helping a country to build national infrastructure for nuclear science and technology programmes and in particular for nuclear power infrastructure. The RR related infrastructure includes the guidelines, laws, regulations, international agreements, education and training support, and support organisations such as an independent regulatory body and/or supporting/oversight government ministries, as well as physical arrangements such as facility, fuel and waste management, radiation protection, safeguards, etc. The specific RR infrastructure is described in the IAEA document Specific Considerations and Milestones for a Research Reactor project, NES No. NP-T-5.1, 2012, which is briefly presented in this paper, including the phases of the development of the RR related infrastructure and the 19 infrastructure issues and conditions to be fulfilled for each of them at the end of each phases. One important question is: Which was the historical role of the RR in developing nuclear power programs and what will be its benefit in the future?, and the paper try to answer to this question based on the experiences of the countries with developed nuclear power program and the newcomers. In the past, the medium to high power RRs were used on more complex research and development activities related to NPP technologies such as fuel and material testing. Also the lower powers RRs, due to their more flexible operation, were more often applied to education -and in less common cases, training- missions. In small nuclear power programme countries there are many examples of utilities bringing licensed NPP operators to RRs for practical/ refresher training and facilitating a better understanding of the laws of physics but this contribution does not comprise a significant portion of the RR’s utilisation schedule. In all cases, these RRs also serve non-power missions related to science, medical and industrial isotope production, and/or other quasi-commercial activities. The majority of newcomer countries with RRs indicated that a RR helped them to be more confident about nuclear power technology and also provided a tool for public information and stakeholder involvement. All the countries embarking on a nuclear power programme are using or have plans to use RRs (although not necessarily their own) for basic education in the nuclear power field (through technical courses and utilisation of the RR) in cooperation with Universities. Most also intend to develop the RR organisation into a future Technical Support Organization (TSO), based on the specific training to be received from the first NPP Vendor country. These information are based on the participation to the different IAEA Integrated Nuclear Infrastructure Review (INIR) missions in newcomer countries. The RRs that are operated in compliance with international standards of nuclear safety, security and safeguards demonstrate a nation’s commitment towards development of nuclear technology in all its aspects.

Speaker: Dr IOAN ROTARU (Nuclear Project Management)

10:10 Design of Miniature Neutron Source Reactor with LEU Core

MNSR with HEU core is mainly used for the neutron activation analysis (NAA), some short-live isotopes production and as training and teaching tool. In order to enlarge the usage of MNSR, such as, medical treatment, physics experiment, prompt gamma neutron activation analysis (PGNAA) and neutron radiography, the neutron beams on the horizontal direction are designed on the both sides of the reactor core. Improved MNSR with thermal power 30kW is an undermoderated reactor of pool-tank type, UO2 with enrichment of 13.0% as fuel, light water as coolant and moderator, and metal beryllium as reflector. In the previous paper, the design of the thermal neutron beam was introduced. This paper will introduce the design of the epithermal neutron beam on MNSR. MCNP code is employed to perform the calculations of the frame design of the epithermal neutron beam, which is mainly composed of neutron moderation layer, thermal neutron absorption layer, gamma ray shielding layer, neutron collimator parts. The moderator materials used probably for the design of the epithermal neutron beam, such as, water, graphite, aluminum, Al2O3, Fluental and other material, are selected according to neutron absorption cross section and scattering cross section of the different nuclides, Fluental with thickness of 50cm is selected as the final moderator material according to the calculating values of the epithermal neutron flux density, gamma and fast neutron contamination at the exit of the epithermal neutron beam for the different moderator materials. Fig.1 [see attached file] shows the diagram of the final design of the epithermal neutron beam, the final calculating parameters at the exit of the epithermal neutron beam are listed in table 1, the results show that the parameters can meet the requirement for Boron Neutron Capture therapy (BNCT)

Speakers: Mr Jin Lu (China Institute of Atomic Energy), Mr Yiguo Li (China Institute of Atomic Energy)

Paper cn231fullpaper.doc cn231fullpaper.pdf

10:30 → 11:00 Break: Poster Session and Coffee Break

11:00 → 12:30 New Research Reactor Project

Convener: Mr Gilles Bignan (CEA)

11:00 KEY TECHNICAL AND SAFETY REQUIREMENTS FOR A NEW RESEARCH REACTOR: THE RA-10 REACTOR EXPERIENCE

A new multipurpose research reactor which will replace RA-3 reactor has been decided to be built in Argentina to satisfy the increasing national and regional demands for radioisotopes. The project, supported by the National Administration, started in 2010 and is planned to be operative in 2019. The expertise acquired in the country, in the design and licensing of nuclear reactors, encourages the National Atomic Energy Commission (CNEA) to face the challenge. INVAP S.E. is involved in the design of the reactor facility and related installations, and will be contracted for the supplying and assembly of SSCs.. Technical and safety requirements have been early established, based on the safety objectives, concepts and principles presented in the NS-R-4. Technical requirements are based on the utilization related requirements and in a consistent reactor availability for the intended use. Key requirements, considering as a guide, the structure and the contents of the IAEA Nuclear Energy Series No NP-T-5.6, have been met as follow: 1. Licensing The Operating Organization (CNEA) developed a Licensing Plan in order to achieve the licensing conditions for the facility. Among other features, this plan includes all the regulatory documentation that must be met in the reactor design process. In the frame of a Design Procedure that includes a Design Plan for each working structure, the corresponding documentation is included as design requirements that must be explicitly fulfilled. In this way the “Design Evaluation” included in the PSAR could be easily performed and concluded. Safety Analysis and the PSAR were elaborated by the Operating Organization. 2. Safety and Radiological Design Requirements The general nuclear safety objective for the RA-10 Reactor is to protect individuals, society and environment from harm by establishing and maintaining in nuclear installations effective defenses against radiological hazards. In order to achieve the safety objective, the project defines three main principles in the design: 2.1 Defense in depth The application of the principle of defense in depth in the design of the research reactor provides a series of level of defense (provision of successive physical barriers, conservative design margins, inherent features, equipment and procedures, including on-site and off-site emergency procedures) which are aimed at preventing accidents and ensuring appropriate protection in the event that prevention fails. 2.2 Safety Functions Safety functions are the essential characteristics functions associated with SSCs that ensure the safety of the reactor. Beside the three fundamental safety function: shutting down the reactor, cooling the reactor core components continuously and confining radioactive material inside the installation, safety functions related to: fission products barrier and configurations integrity, control and monitoring, reactivity regulation and shutdown, reactor protection, radiative material confinement, core cooling, services supplying, shielding, decayment and purification and physical protection were defined and classified based on their importance to safety. Safety functions were identified to ensure the effectiveness of each level of defense in depth. 2.3 Acceptance criteria and design rules Acceptance criteria were established for operational states, for DBAs and for selected BDBAs. For operational states, radiation exposure within the installation or due to any planned release of radioactive material from the installation is kept below prescribed limits and as low as reasonably achievable For DBAs and BDBAs, all accident sequences were evaluated and demonstrated to comply with the regulatory established acceptance criteria, i.e. their annual occurrence probability combined with the corresponding effective dose must meet a limiting risk level according to the national regulations. Engineering safety features for DBAs and for BDBAs are implemented to meet this criterion. For the design of SSCs, acceptance criteria are established in the form of engineering design rules. These rules include requirements related to the classification of the SSCs that are important to safety. The operative experience of the CNEA and the design experience of INVAP were the main tools for meeting the acceptance criteria. 3. Utilization related design requirements Utilization related requirements provided by the user were consolidated and restricted, resulting in the following utilization related design features: neutron activation analysis, radioisotopes production, neutron transmutation doping, neutron beam applications for material structure studies and neutron radiography. 4. Others key technical requirements The availability of the reactor is required to be consistent with its intended use. In order to assure an availability of 80% for the reactor and of 90% for the facilities, availability factors were assigned to SSCs considering its need for the reactor or facilities operation. These availability factors resulted in a quality classification for all ESCs. The spent fuel storage facility is required to provide space for the volume of ten years operation (20 cores). The facility will be placed in the Reactor Service Pool and will be compatible with a transportation cask for carrying the spent fuel to a general storage facility.

Speaker: Mr Herman Blaumann (Comisión Nacional de Energía Atómica)

Paper IAEA-CN-231-74.pdf

11:30 The RMB Project - development status and lessons learned.

The RMB Project - development status and lessons learned. PERROTTA, J. A., OBADIA, I. J., SOARES, A. J. Comissão Nacional de Energia Nuclear (CNEN), Av. Prof. Lineu Prestes 2242, Cidade Universitária, 05508-000, São Paulo, SP, Brazil. Corresponding author: perrotta@ipen.br Brazilian research reactors and related facilities have a limited capacity for radioisotope production, leading to a high dependence on external supply for radioisotopes used in nuclear medicine. In order to overcome this condition and due to the very old age of the main research reactors in the country, the Brazilian Nuclear Energy Commission (CNEN) decided to start a new research reactor project, named RMB (Brazilian Multipurpose Reactor). This reactor will be part of a new nuclear research center, to be built on a site about 100 kilometers from São Paulo city, in the southeast part of Brazil. The new nuclear research center will have a 30 MW open pool type research reactor using low enriched uranium fuel, and several associated facilities and laboratories in order to produce radioisotopes for medical and industrial use; to use thermal and cold neutron beams in scientific and technological research; to perform neutron activation analysis; and to perform materials and fuel irradiation tests. This article presents updated information on technical issues as well as on the overall development status of the RMB project. It also discusses some technical matters and lessons learned related to the complexity of the project management.

Speaker: Mr jose augusto perrotta (CNEN)

Paper ID_32.pdf

12:10 MBIR INTERNATIONAL RESEARCH CENTER. CURRENT STATUS AND PROSPECTS

The Multipurpose Sodium Fast Research Reactor MBIR is under construction at the site of SSC RIAR, Dimitrovgrad, Ulyanovsk region in the framework of the Russian Government’s funded Federal Target Program «New Generation Nuclear Power in 2010-2015’s and up to 2020». The MBIR Reactor which would be the most powerful research facility amongst the group of research reactors both operative, under construction or planned worldwide is intended to replace the world’s unique fast research reactor BOR-60, which was built about half a century ago and has been safely operating all these years at the SSC RIAR’s site [ 1 ]. The top intended mission of the MBIR Reactor is the significant expansion of the ROSATOM’s experimental capabilities and the overall enhancement of international R&D infrastructure for future innovative nuclear power system development. The MBIR Reactor with the unique technical parameters and consumer properties is aimed at solving of the wide range of actual research tasks toward long-term nuclear technology and closed nuclear fuel cycle’s justification and demonstration. Besides that, the multitask design and technological features of the MBIR Reactor provide MBIR’s using for the applied issues and technology development in the field of medical applications, space and electronic industries, production of radioisotopes and radiation-modified materials. Basic parameters of the MBIR Research Complex is the following [ 2, 3 ]:  thermal power 150 MW  maximum neutron flux 5,3•1015 n/(cm2•s)  designed life time 50 years  upgradeable experimental capabilities: up to 3 External Loop Channel with different coolant type (Na / Pb / Pb-Bi / He etc.) up 14 Material Test Assemblies, MTAs with different design options in the core (plus up to 36 MTAs in 1st row of blanket) up to 3 Instrumented In-Pile Experimental Devices (design is developed by a separate performance specifications depending on the User’s experimental tasks) horizontal and vertical experimental channels for neutron radiography, physical researches, medical applications and Silicon’s radiation-doping  wide range of post-irradiation examination, PIE facilities and additional facilities (experiment preparation & support labs, analytical labs with the cutting-edge examination devices, offices & meeting rooms etc.) in the MBIR Research Complex

Speaker: Mr Alexander Tuzov (Joint Stock Company "State Scientific Center - Research Institute of Atomic Reactors")

Paper MBIR_International_Research_Center.pdf

12:30 → 13:30 Lunch Break

13:30 → 15:00 Spent Fuel Management and Decommissioning

Convener: Mr Yidong ZHOU

13:30 MANAGEMENT OF TRANSITION BETWEEN SHUT DOWN AND DECOMMISSIONING OF RESEARCH REACTOR CIRUS

Cirus, a vertical tank type 40 MW thermal research reactor, natural uranium fueled, heavy water moderated and light water cooled; is located in Bhabha atomic research centre, Mumbai, India. It achieved first criticality in July 1960 and was operated successfully for 50 years till permanent shut down in December, 2010. The core has been completely unloaded, heavy water has been removed and reactor systems have been brought to a safe state in preservation mode with minimum surveillance requirement to conserve manpower and save energy and effort. A deferred dismantling (safe enclosure) has been chosen as the decommissioning strategy for the reactor. A detailed plan has been prepared for managing the transition between permanent shut down and deferred decommissioning and executed. Some of the jobs in progress include radiation mapping of reactor structure, estimation of radioactivity content and decay pattern; introduction of technical specifications and surveillance methodology, estimation of waste generation and its characterization; categorization of components for reuse in other facilities, release for unrestricted use and or scrap; Sampling for data generation on irradiation and corrosion damage suffered by materials, development of decontamination techniques, etc. A preliminary decommissioning plan has been prepared. The presentation will cover all aspects of managing the transition and highlight experience gained in post permanent shut down management of Cirus.

Speaker: Mr Ram Sharma (BARC)

Paper Final_Paper-India.docx Final_Paper-India.pdf

Slides 38_Cirus_-_transition_from_operation_to_decommissioning,___modified_for_the_paper.ppt

14:00 Safety assessment of the Osiris Research Reactor and final shutdown officering

The French research reactor “Osiris” is an open-core pool type reactor notably used for the irradiation of enriched uranium targets in order to produce medical isotopes and operated by the French Alternative Energies and Atomic Energy Commission (CEA). Designed in 1965, it was supposed to be definitively shut down by the end of 2015. This assumption has been considered by the French Nuclear Safety Authority (ASN) and the Institute of Radioprotection and Nuclear Safety (IRSN) during the partial safety reassessment carried out by the CEA in 2009: several safety topics either were not evaluated or were evaluated with lower safety requirements. At the end of 2011, the CEA asked the French Prime Minister for the authorization to keep the reactor operational until 2018. The CEA motivated its request by the potential lack of medical isotopes during the period between the Osiris reactor final shutdown and the Jules Horowitz reactor (RJH) commissioning (the RJH is a new pool type research reactor under construction in Cadarache - France). In order to assess this demand, the ASN needed to determine whether the Osiris reactor met the current standards, as other perennial reactors. So the ASN asked the IRSN to initiate a safety assessment with the aim of identifying the main modifications that would be necessary to ensure that the reactor is operating with sufficient safety conditions for a significant time. The paper will present the reactor safety assessment performed by the IRSN in 2013 in order to identify the parts of the reactor’s safety case that are consistent with the standard nuclear engineering practice and to point out the aspects of the safety case that should be reassessed if the operation were continued beyond 2015. In particular, the paper will present the opinion of the IRSN about the approach followed by the CEA to demonstrate the reactor’s safety. Indeed, if the reactor had to be operated longer, a safety demonstration based on operating conditions would have been required. Such an approach actually structures the safety demonstration and enables to better judge the provisions implemented for the different levels of the defense in depth. The paper will also present the assessment of the reactor containment performed by the Institute, which led to the conclusion that the CEA would have to reinforce the static containment integrity in order to continue operating. Indeed, the total radioactivity release in environment due to a BORAX-type accident (namely radioactive Iodine and noble gas release) could be significantly reduced if Osiris building reactor’s gas-tightness were improved. A reinforcement of the containment building would indeed allow a different handling of the post-accidental situation. The paper will present the way the ASN has officered the final shutdown of the facility considering safety aspects. It will also present the work done by the ASN regarding medical radioisotopes supply in the past years. The paper will finally present the position taken by the ASN in 2014 regarding the request of the CEA to keep the reactor operated after 2015.

Speakers: Mr Christophe BIGOU (ASN), Mrs Stephanie KANAMORI (IRSN)

Paper ID_35.pdf

14:20 Decommissioning of Georgian Nuclear Research Reactor

1. Introduction Georgia had only one nuclear research reactor IRT-M belonged to the Institute of Physics. The pool type reactor was put in operation at 1960 and shut at 1988. During its operation the reactor was refurbished two times due to what its power was increased from 2 MW up to 8MW. Accordingly, the neutron flux was increased up to 1014n/cm2•sec. The especial investigations were conducted to research properties of different materials exposed by neutrons on low temperatures (temperature of liquid nitrogen and helium), therefore the reactor had special station and circuits for nitrogen and helium. For 28 years (1960-1987) the nuclear reactor was operated more than 70,000hours. More than 9 000MW•day heat energy was produced, which corresponds to the consumption of nuclear fuel - Uranium-235 in the amount of 11 kg. The reactor had used various types 201 fuel assemblies, in which the total content of 90% enriched Uranium- 235 was about 30kg 2. Preliminary Decommissioning Activity At 1990 considering real situation: absence of storage and processing facilities, technique, affordable finances and skills for conducting the full scale decommissioning activity (which fully corresponds the criteria [1]), the decision was issued to conduct the differed decommissioning. The decommissioning plan DDP was developed aiming transfer of the reactor to the low power nuclear facility. According to the plan, after obtaining of special permissions, fresh fuel was sent to the same type reactor and spent fuel for refurbishing. The plan considers conducting of concreting of lower part of the reactor tank where the reactor core is situated. The special Barium containing concrete is supposed to use to cover reactor core, some high active reactor parts (especially placed there) and experimental channels [2]. Before the concreting the calculations were conducted to assess the exposure on the concrete top using the special software. (Fig.1). According to the plan the following activities were conducted: 1. Preparation of the reactor hall; 2. Preparation of the reactor tank with its internals; 3. Preparation of the experimental horizontal channels; 4. Preparation of the radioactive waste; 5. Preparation of the auxiliary systems and reactor equipment; 6. Concreting the reactor tank; 7. Concreting the 8 experimental horizontal channels; 8. Concreting the waste in the dry storage vertical channels; 9. Concreting the waste in the storage well; 10. Installation of the monitoring and surveillance system. All activities were conducted within IAEA TC Project GEO/4/002 “Conversion of Research Reactor to a Low Power Facility”. To provide safety conditions during the concreting activity the special underwater concreting was conducted (Fig.2). The measured data has good compliance with calculated ones. 3. Decommissioning of Axillary System The decommissioning of axillary systems was started only after putting into operation of Centralized Storage Facility. All systems in the reactor hall was dismantled within IAEA project GEO/3/002. Considering IAEA standards [3] the special decommissioning plan was elaborated and approved by regulatory Body. The plan considers existence of indicators for decommissioning activity [4]. According the plan the following systems should be dismantled: 1) The dual-circuit cooling system of the reactor. 2) The system of mechanical and chemical purification of the coolant of the primary circuit of the reactor cooling system. 3) The part of the pipeline of the system of circulation of gaseous helium. 4) The system of filters intended for cleaning the air from radioactive gases and aerosols being ventilated from the above-reactor space and different special technological rooms prior to their release into the atmosphere. 5) Devices of mechanical and chemical purification of water of pools intended for temporary storage of the fuel assemblies and cassettes. The preliminary radiation monitoring fixed 60Co and 137Cs as main contaminant radionuclides. The secondary cooling circuit was assigned as a radiologically clean (under clearance level). The average contamination value by 60Co was 20-30 Bq/cm2 and dose rates near the contaminated surfaces were in the range 5-50 mSv/h. Not having for this time the capability to clean safely all dismantled parts, the decision was issued to close hermetically all contaminated tubes and put into the storage facility. Only some small parts with comparably high activity (including resigns) were immobilized into the concrete drums 926 drums). The same technology was used to dismantle huge pipelines connected the reactor hall to the cryogenic station (The activity was conducted under IAEA TC project GEO/3/004). The total activity of the pipes under dismantling was 1.8x109 Bq. The pipe cleanings started with IAEA TC project GEO/9/011. The Cryogenic station will be dismantled within IAEA Project GEO/9/012 REFERENCE 1. IAEA WS-G-2.1 2. IAEA TECDOC 1124 3. IAEA SRS-45 4. IAEA NW-T-2.1

Speaker: Prof. Giorgi Nabakhtiani (Departent for Nuclear and Radiation Safety)

14:40 Coatings for safe long term wet storage of spent Al-clad research reactor fuels

Spent aluminum-clad fuels from research reactors (RRs) are stored for decades in water filled pools. Pitting corrosion of these fuels has been reported despite water quality management programs at the storage sites, and attributed to synergism in the effect of specific basin water parameters on corrosion of aluminum alloys. Corrosion protection of the spent RR fuel in wet storage is therefore considered to be important. Coatings are widely used to protect metallic materials in many industries. The shape of the RR fuel and radioactivity of spent fuels especially preclude coatings formed electrochemically. Hence, chemical surface treatment to form a protective coating is the only option and in this context conversion coatings were considered. The results of preliminary laboratory and field investigations carried out at IPEN in Brazil revealed that chemically formed rare earth element containing oxide coatings increased the corrosion resistance of Al alloys. These investigations were extended to include boehmite (a form of hydrated aluminium oxide) and hydrotalcite (HTC) coatings, which is lithium aluminium-nitrate-hydroxide hydrate that forms on Al alloys immersed in an appropriate alkaline lithium salt solution. The objective of this investigation was to: (a) prepare and characterize hydrotalcite coatings from baths at different temperatures followed by post-coating treatments such as cerium oxide incorporation and sealing; (b) determine from laboratory tests the corrosion behavior of HTC coated AA 6061 alloy in NaCl; (c) evaluate the extent of corrosion protection offered by the coatings from field studies in which uncoated and coated AA 6061 alloy coupons and ‘dummy’ fuel elements were exposed to the IEA-R1 reactor spent fuel basin for periods of up to 24 months. The HTC coating preparation procedure consisted of immersion of AA 6061 alloy specimens for the laboratory tests and coupons as well as full-size plates for the field tests in solutions to: (a) degrease; (b) deoxidize; (c) form the HTC coating; (d) incorporate cerium oxide in the coating; (e) seal the coating. The HTC coating consisted of intersecting blade-like crystallites and that formed at 98 °C was significantly thicker compared to the coating formed at room temperature (RT). The corrosion behaviour of uncoated and HTC coated specimens was determined from anodic potentiodynamic polarization measurements in 0.01 M NaCl. These measurements indicated marked increase in the corrosion resistance of the HTC coated specimens. Further, the effects of duration of treatment in the coating solution and the cerium solution were also determined. The field tests consisted of: (a) preparing uncoated and coated coupons and plates; (b) stacking of the coupons in racks and assembling of the plates to form full-size 'dummy' fuel elements; (d) immersion of the racks and the ‘dummy’ fuel elements in the spent fuel section of the IEA-R1 reactor in IPEN, Brazil, for periods of up to 24 months; (e) removal of the racks or ‘dummy’ fuel elements, rinsing and decontamination; (f) dis-assembly and examination of the coupons and plates. The room temperature or RT-HTC coated coupons revealed a few pits after 5 months of exposure whereas the HT-HTC coated coupons exposed for the same duration did not reveal any pits. On the basis of these observations the full-size plates were coated with only HT-HTC and not RT-HTC. The HT-HTC coated plates from the ‘dummy’ fuel elements exposed for 14 months were stained but revealed no pits. The cerium incorporated HTC imparted much higher corrosion protection. Sealing of the coatings further improved the pitting corrosion resistance of the alloy. The HTC layer imparts pitting corrosion protection by acting as a physical barrier between the solution and the surface. The mechanism by which the cerium in the HTC imparts protection is considered to be ‘active corrosion protection’. According to this mechanism, the lower solubility of CeO2.2H2O allows formation of Ce4+ ions in solution which then diffuse to defects in the coating and exposed bare metal. When in contact with the bare metal, these ions reduce to Ce3+ and precipitate as Ce(OH)3 and thus seal the defect in the coating.. The coating procedure for irradiated Al-clad spent fuels would be facilitated if all the treatments were to be carried out at room temperature. Currently studies are in progress to pre-treat, coat and post-treat the Al alloys to obtain highly protective HTC coatings at room temperature. Details of these studies and the extent of corrosion protection offered by this coating will be presented and discussed.

Speaker: Dr Lalgudi Ramanathan (Instituto de Pesquisas Energéticas e Nucleares - IPEN)

Paper IAEA_-_RR_conf_Nov_16-20_2015.pdf

Slides 57_Ramanathan.pptx

15:00 → 16:00 Break: Poster Session and Coffee Break

16:00 → 17:50 Security of Research Reactors

Convener: Mr Faizan Mansoor (Pakistan Nuclear Regulatory Authority)

16:00 Risk– Based Approach for Security Management

Nuclear security addresses two specific global concerns: (1) the initiation of unacceptable radiological consequences through the intentional, malicious dispersal of nuclear and/or other radioactive materials; and (2) the theft of nuclear and/or other radioactive materials with the intent to construct a weapon of mass destruction. Thus, nuclear security provides, at once, a vital complement to both nuclear safety and nuclear safeguards. Nuclear security is a necessary component of these global programs to ensure that the benefits of nuclear and radioactive materials are preserved for society by protecting against adversary-induced unacceptable consequences. Currently, the international community places significant emphasis on a robust nuclear security regime. This can be seen by the large number of international instruments highlighting the State’s obligations and responsibilities with respect to nuclear security; the availability of international conferences and training courses focused on nuclear security; the popularity of IAEA nuclear security missions and guidance documents; and the number of donor States investing significant capital and resources to assist global strengthening of nuclear security. For those not intimately involved with nuclear security, the basis for this emphasis may not be obvious. This presentation will provide a background of what led to this emphasis and describe how nuclear security is implemented to help ensure that the societal benefits of nuclear and radioactive materials are preserved. The euphoria that accompanied the “Atoms for Peace” initiative of the 1950’s and early 60’s obscured any nuclear security considerations. Research reactors were designed and constructed to optimize their intended operations. However, beginning in the 70’s, concerns about security began to arise due to changes in the global threat environment. In 1972, in response to activities of several militant groups around the world and highlighted by the events of the Munich Olympics, the Director General of the IAEA invited security experts from a few Member States to develop the first international security recommendations. The document, “Physical Protection of Nuclear Material” was soon published by the IAEA. This was the first international nuclear security document, but it only dealt with theft of nuclear material (and hence, proliferation concerns). Not long afterward, a revised and more inclusive security document, INFCIRC 225, was published (1975). This was followed by the development of the International Convention on the Physical Protection of Nuclear Materials and Nuclear Facilities in 1980. Subsequent events, such as truck bomb attacks to buildings, suicide bombers, and nuclear material trafficking events in the early 1990’s, led to steady increases in the attention given to nuclear security and in particular to DBTs, insiders, and sabotage. The attacks of September 11, 2001 prompted the development of several international nuclear security instruments, an even greater emphasis on nuclear security at the IAEA, concerns of security for radioactive materials, and greatly increased support from donor States for security assistance. Paralleling these events and their associated response by the international community, the concepts and approaches to nuclear security were steadily improved. The result of this steady improvement is a mature, structured, and systematic approach to nuclear security. This approach includes the security responsibilities and coordination of both State bodies and operator organizations under a nuclear security regime. It considers the contribution of technical and administrative measures to achieve the fundamental security functions of detection, delay, and response in an integrated and balanced manner that serves to both deter and prevent theft and sabotage. It advocates a graded-approach philosophy, whereby more attractive targets are afforded more robust security. It provides a validated performance-based methodology to enable operators and State authorities to assess the effectiveness of the security system against credible adversary threats, thereby providing confidence that the security system is adequate. This systematic, performance-based approach lends itself to establishing risk-informed security levels for a research reactor. The potential security risks posed by the research reactor facility can be assessed by understanding: (1) the potential radiological consequences of intentional, malicious acts, (2) the ease with which the consequences can be intentionally initiated, and (3) the safety system effectiveness in mitigating these consequences, and (4) the security system effectiveness in preventing or complicating the ease of initiation. Once understood, the risks can be “managed” by increasing or decreasing the nuclear security system effectiveness. By modifying the features of the nuclear security system and measuring the changes in estimated risk, a nuclear security system can be identified that optimizes the many parameters that impact the “ideal” security system. These parameters are: (1) the risk posed by security threats, (2) the costs of installing and operating a nuclear security system to adequately mitigate these risks, and (3) the operational impacts of specific security measures. This risk informed approach to nuclear security has evolved over this period of increasing threats, and has been adopted by the international community.

Speaker: David Ek (Sandia National laboratories)

Paper 157_Risk_based_security_management_at_research_reactors__paper_G01.pdf

Slides 157_Risk-Based-Approach-Security-Mgmt-FINAL.pptx

16:30 Security Management of Research Reactors and Associated Facilities

The title “RRAF” represents a diverse category of non-power reactors that can include a wide variety of co-located facilities that can complicate a security system. These facilities may include: • Nuclear reactor • Radioisotope production facilities • Fuel research and fabrication facilities • Storage of fresh fuel, spent fuel, or radioactive sources • Radioactive waste storage and disposal • Laboratories and hot cells • Irradiation facilities • Other non-nuclear related facilities and activities RRAFs, due to their diverse objectives, settings, funding, and staffing present a unique set of challenges to the implementation and maintenance of an effective nuclear security programme and management. These diverse objectives for the security system in the RRAF include drivers such as country-specific security demand for "critical" infrastructure, security demands coming from the need to protect assets other than the nuclear and radiological materials, and their associated facilities and activities in the research centre or from the need to mitigate other than radiological consequences, such as reputation damage or disadvantageous political consequences such as losing public opinion support for the activities in case a security event would happen and create a perception that the security system was incomplete, imperfect or deficient. The overall objective of a State’s nuclear security regime (as per NSS#13 and NSS#14) is to protect persons, property, society, and the environment from malicious acts involving nuclear and other radioactive materials. In line with the objectives of the IAEA Nuclear Security Series, the objective of the Technical Document, Security Management for Research Reactor Operators is to develop a comprehensive, systematic, and focused approach to assist research-reactor operators in Member States to establish, enhance, and sustain appropriate nuclear security management at RRAF sites. Nuclear security management is a term used to describe the roles and responsibilities that must be effectively performed and the programmes or functions that must be successfully implemented in order that the RRAF nuclear security programme meets the objectives laid out in the international instruments, IAEA recommendations and guidance, and the State’s regulatory requirements. RRAF reactor facility management require operator-centric guidance that focuses on assisting operators to implement an effective security management system. The TecDoc provides guidance to management in demonstrating the effectiveness of their security programme to the Competent Authority. The Competent Authority may also find the TecDoc useful in the licensing and inspection of the operator’s nuclear security system. A RRAF Nuclear Security Management System (NSMS) includes roles, responsibilities and programmes for management of security at a facility in three topical areas: Security operations, Security Processes and Security Forces, and its relationship with the State’s nuclear security regime. There are also many security related activities, relationships as well as governance and organizational interfaces that drive a NSMS i.e. legislation, regulations, processes, and plans that will enable a State to effectively regulate the use, storage, and processes of nuclear and radioactive materials and facilities. At any facility, there should be an overall organizational Facility Integrated Management System (FIMS) that integrates all the organization's systems and processes into one complete framework, enabling an organization to work as a single unit with unified objectives. An integrated system provides a clear, holistic picture of all aspects of the organization as well as how they affect each other and their associated risks. A FIMS allows a management team to create one overall structure that can help to effectively and efficiently deliver an organization's objectives. It is important that the management structure for nuclear security at a RRAF be clearly defined. It is the facility nuclear security management that must define roles, responsibilities, and accountabilities for each level of the organization, including security and other interfaces. This would also include an explicit, specific assignment of accountability for nuclear security to an individual (or individuals) along with the necessary authorities, autonomy, and resources to successfully implement this role. The individual(s) must report to the top manager or appropriate senior manager of the organization and the responsibilities of their position must be defined/documented in sufficient detail to prevent ambiguity. Managers are responsible for ensuring that appropriate standards of behaviour and performance associated with security are set and that expectations as to the application of these standards are well understood. They must also ensure that there is a clear understanding within the organization of the roles and responsibilities of each individual involved in security, including clarity concerning levels of authority and lines of communication. Managers should also establish a formal decision-making mechanism that is well understood within the organization and involve their staff in decision making processes, where appropriate. The quality of a decision is improved when the individuals involved are able to contribute their insights, ideas, and experiences. The security system must be in a continuous state of readiness in order to handle security events at any time.

Speaker: Mr Eric Ryan (Nuclear Security Consultant)

16:50 SAFETY CONSIDERATIONS WHEN IMPLEMENTING SECURITY AT RRAFs

Concern about malicious acts involving Research Reactors and Associated Facilities (RRAFs) is not new; however, recent world events have demonstrated that an attack on RRAFs must be strongly considered and protected against. Changes in the global threat environment have served as an incentive for RRAF’s organizations to give adequate attention to nuclear security of their facilities. Many RRAF organizations are currently planning or implementing nuclear security measures. In parallel to these efforts, the majority of RRAF organizations continue to implement modification and refurbishment projects as mitigating measures for ageing facilities. In addition, more than 20 Member States are currently implementing (or planning) new RRAF facilities. Most of these Member States are planning their first RRAFs and are in a process to develop the necessary nuclear safety and security infrastructures. Although the activities addressing nuclear safety and nuclear security focus on different protection aspects, they can sometimes overlap each other. Actions that are taken to further one activity can have implications on the other. Recognizing that nuclear safety and nuclear security share the same ultimate goal–to protect individuals, the public, and the environment from harmful effects of radiations–a well-coordinated approach is mutually beneficial, and safety and security measures should be established and implemented in a manner that they do not compromise one another but rather they should mutually enhance. The interface between nuclear safety and nuclear security should be addressed in an integrated manner throughout the RRAF lifetime. This paper provides information for a better understanding that nuclear security management at RRAFs could have an impact on nuclear safety, so that any security action should be fully analysed and its interface with safety understood to mitigate potential negative impacts.

Speaker: Prof. Jose Lolich (Instituto Balseiro)

17:10 International Policies and Tools for Protecting Against Radiological Sabotage

This paper will provide a comprehensive overview of international policies and tools that can be used for protecting against radiological sabotage at nuclear and radiological facilities. This analysis is uniquely relevant for international nuclear security because there are few materials available which provide a comprehensive overview of both policy and technical context and issues associated with sabotage. The paper will be geared toward policy makers and managers who do not necessarily have a strong understanding or familiarity with the concept of sabotage and the roles and responsibilities of various stakeholders. Through describing the rising threat of non-state and unconventional actors, we will provide historical context for the threat of nuclear terrorism and sabotage After providing historical context for the rising threat of sabotage, we will describe several international policies and guidance documents geared toward addressing the threat of sabotage, including Nuclear Security Recommendations on Physical Protection of Nuclear Material and Nuclear Facilities (INFCIRC/225/Revision 5) and the Convention on the Physical Protection of Nuclear Material (CPPNM). We will describe and provide an analysis of the factors which influence the scale of sabotage threat, including material type and quantity, type of radioactive material release (single or sustained), and type of sabotage scenarios. Several potential sabotage scenarios which can be identified in a threat assessment or design basis threat include direct versus indirect attack and a blended cyber-physical attack. Through highlighting the roles and responsibilities for the Competent Authority and site operator, the paper illustrates how respective authorities conduct threat assessments, determine thresholds for unacceptable and high radiological consequences, and identify vital areas within a site or facility which may warrant further protection. In addition, we analyze how regulators and operators can cooperate to implement mitigation measures or contingency plans in response to various sabotage scenarios. Finally, the paper will review several software tools used to analyze sabotage threats and consequences, including the U.S. Nuclear Regulatory Commission’s Radiological Assessment System for Consequence Analysis (RASCAL), Sandia National Laboratory’s Turbo Federal Radiological Monitoring and Assessment Center (FRMAC) software, the National Atmospheric Release Advisory Center’s (NARAC) HotSpot Health Physics code, and the Quick Look Radiological Assessment Methodology (QLRAM) developed by Pacific Northwest National Laboratory for the National Nuclear Security Administration.

Speaker: Mr Rosalyn Leitch (PNNL)

Paper IAEA_RR_FINAL_Paper_166_-_Leitch_-_International_Policies_for_Sabotage.docx

Slides IAEA_RR_Final_PPT-Paper_166-Leitch.pptx

17:30 Methodology on Cyber Security for Digital I&C system in Research Reactors

Cyber security has become one of major issues in nuclear field both commercial power plants and research reactors. For cyber security in nuclear field, the regulatory agencies in the world published a lot of guidance. It is necessary to evaluate cyber security considering the conformance with these regulatory guides. In this study, we introduce the cyber security risk evaluation model with Bayesian network (BN) for cyber security of digital instrumentation and control (I&C) system in particularly, research reactors. We propose the methodology of making the event tree which is one of tools for probabilistic safety assessment (PSA) for risk management. Our first sample system was the nuclear reactor protection system (RPS), one of the safety-critical systems, for cyber security risk evaluation. The BN has an advantage of easily modeling complex dependencies, and being useful for composing the model which uses prior information and posterior information, and back propagation calculation. The BN is used for vulnerability study of cyber security, for which it is difficult to perform the penetration test. For a nuclear facility which a penetration test cannot be conducted in reality, the quantitative value provided by the BN model can address cyber security in research reactor with quantitative manner. The proposed cyber security risk evaluation model consists of two views: the activity-quality model and the architecture model. The activity-quality model analyses how people and/or organization comply with the cyber security regulatory guides for nuclear facilities. Meanwhile, the architecture model analyses vulnerabilities and mitigation measures according to architectural characteristics of RPS on the basis of the BN. The integrated BN model can be used to evaluate comprehensive cyber security risk for RPS. Furthermore, it informs the useful information about vulnerability risk by using input branch values in the event tree model. Though the node probability tables were decided by expert judgment as prior information, we will perform Bayesian update to get posterior information for using true data by experiment test to make more robust the model. The conventional PSA model does not consider the cyber security yet. However, it is expected to cover the effects of cyber-attacks in risk management. In order to achieve this technical goal, we tried to use event tree models, which enables to represent the cause-consequence relationships between vulnerabilities and mitigation measures of an I&C system. In terms of cyber security, vulnerability can be considered as an initiating event in an event tree model since the vulnerabilities are the starting point of cyber-attack. The input value of the branch probability in the event tree model was referred from the analysis results obtained from the BN cyber security risk evaluation model. The event tree model is also able to show the visual information about the mitigation measures for each vulnerability with prior information, posterior information, and back propagation calculation. Furthermore, it informs the quantitative information about vulnerability risk based on the input branch value from the BN model. After Bayesian update for the BN model with experimental data, it provides the vulnerable route on target architecture by evaluating the cyber security risk instead of penetrate test. The cyber security risk evaluation model can substitute for enormous penetration test on nuclear facility by evaluation the vulnerability against cyber-attack and providing information that which route or point has more vulnerable than others. It can be used to prepare the cyber security analysis report for regulatory authority.

Speaker: Prof. Gyunyoung Heo (Kyung Hee University)

Paper Methodology_on_Cyber_Security_for_Digital_I&C_System_in_Research_Reactors_Submittion.pdf

Slides G05_Heo.pptx

19:00 → 21:00 Conference Dinner

Thursday, 19 November

09:00 → 10:30 Safety of Research Reactors

Convener: Mr David Newland (Canadian Nuclear Safety Commission (CNSC))

09:00 Periodic safety reviews: basis and benefits of improving safety

To ensure the effective fulfillment by operator of national and international safety requirements, the systematic safety reassessments of the Nuclear Research Facility (NRF) should be performed periodically on the basis of graded approach with due account to potential hazard associated with a specific facility. The regulatory body establishes frequency of the safety assessments to be performed by the operating organization, and identifies a set of safety issues to be considered during the assessment. Based on the assessment (self-assessment) results the regulatory body takes decision with regard to the acceptability of the NRF safety and its continued operation for the period before the next planned reviewing. If the safety requirements are not fully met, the operating organization must make actions to reach or restore the required level of safety including NRF modifications. The report presents information on the methodology and experience of application of the Periodic Safety Review (PSR) for enhancement of NRF safety that covers assessment of all aspects of NRF safety operation and involves also the cumulative effect of various factors affecting the assessment results. From the report it is clear that PSR provides a consistent, reliable means for identifying and taking timely preventive measures to eliminate deficiencies in safety of NRF operation and is an effective tool for improving of safety through implementation of international good practices by the both operating organization and regulatory body.

Speaker: Alexander Sapozhnikov (Deputy of Department Head)

Paper Sapozhnikov-PSR-final7.pdf

Slides

• C10_Sapozhnikov.ppt
• PSR__16-20_Nov_2015__Russia_Sapozhnikov_Angl-10.ppt
• PSR__16-20_Nov_2015__Russia_Sapozhnikov_Angl-11.ppt
• PSR__16-20_Nov_2015__Russia_Sapozhnikov_Angl-9.ppt

09:50 The first PSA of the TRIGA Mk II in Vienna, Austria

The concept of a ten year safety assessment (Periodic Safety Assessment, PSA) of a nuclear facility was introduced into Austrian legislation only in 2011, resulting in a deadline for a first such assessment by the end of 2014 for existing facilities. The content of the PSA as laid down in the regulations consists of - The current Safety Report of the facility - Presentation of the current condition of the facility - Analysis of the Operational Experience The current version of the Safety Report of the TRIGA Mk II in Vienna was recently approved by the authorities. As Austrian law already requires a yearly safety assessment for any Nuclear facility, the production of the PSA was in its first iteration a collection of the findings of the last 10 yearly assessments. Because of the yearly assessment by the regulatory bodies, a strict and detailed plan for maintenance and inspection of all relevant systems of the research reactor exists. From this, a detailed description of the current condition of the facility could easily be extracted. Following the structure of the periodic maintenance schedule and in-service inspection, every relevant system and part of the research reactor was presented, any work performed on the system within the last 10 years was described and a conclusion to the current status of the system was drawn. One of the conditions from the last yearly assessment of the TRIGA Mk II research reactor was an exchange of the old I&C system, to bring the instrumentation of the reactor fully up to date. With this in mind, it was decided to use the PSA to exchange or renew most of the major infra-structural components of reactor. As such, many systems that were still functional but close to their expected life time, were marked for exchange in the PSA, most important the primary and secondary cooling system as well as the reactor ventilation system. Because of rigorous records kept by the reactor personnel from the start up of the reactor in 1962, an in-depth analysis of all operational measures could be performed. All notes in the reactor log book were entered into a database and assigned to the categories defined in the IAEA safety standards. Throughout the operational lifetime of the reactor, only 3 events occurred that were subject to report to the authorities. From the chronological analysis of the records, the renewal of the I&C systems was indicated. After submission of the PSA to the competent authorities in late 2014, and a subsequent review of the documents by experts, a few additional documents were required for final submission. The PSA is currently being processed by the regulatory bodies.

Speaker: Dr Johannes Sterba (Atominstitut, Vienna University of Technology)

Paper 117_Sterba_etal_PSA_TRIGA_Vienna_final.docx

10:10 Polish Regulatory Authority experience from INSARR mission for the Maria research reactor

The National Atomic Energy Agency (PAA) in Poland undergone the IAEA Integrated Safety Assessment of Research Reactors (INSARR) mission dedicated for the Maria research reactor during the period from 31 March to 4 April 2014. Maria reactor is located is in the city of Swierk, 30 km from the city of Warsaw. It is a tank-in-pool, multipurpose, cooled and moderated by light water and beryllium with graphite reflectors, with a maximum power of 30 MW. Currently, the reactor is normally operating at a power level between 18 and 23 MW. The reactor has been designed and manufactured by the Polish Industry and went critical for the first time in 1974. The reactor is currently operated with fuel elements with enrichment of 19,75%. The programme of core fuel conversion to low enriched uranium was completed in September 2014. The reactor is expected to play an important role in the development of the national safety and technical infrastructures for embarking on a nuclear power programme. The Regulatory Authority demonstrates the current status of the implementation of the INSARR guidance and its assessment from the mission. Majority of the work will include lessons-learned realization during the License Renewal Process for Maria RR, and evaluation of the Safety Analysis Report in 2015.

Speaker: Ms Justyna Adamczyk (National Atomic Energy Agency)

Paper ID_61.pdf

Slides CN-231-61_Adamczyk.ppt

10:30 → 11:00 Break: Poster Session and Coffee Break

11:00 → 12:30 Safety of Research Reactors

Convener: Mr Walid ZIDAN

11:00 The Application of a Graded Approach in the Regulation of Research and Test Reactors at the U.S. Nuclear Regulatory Commission

The concept of a graded approach is a basic aspect of the International Atomic Energy Agency (IAEA) for maintaining safety at research reactors. A graded approach is also a fundamental aspect of the regulation of research and test reactors (also called non-power reactors) by the U.S. Nuclear Regulatory Commission (NRC). The purpose in applying a graded approach is to match the degree of scrutiny exercised in the regulatory process to the safety significance of the features or characteristics of the design that is being evaluated. Research and test reactors regulated by NRC encompass a multitude of designs and power levels. Thermal power levels and designs range from 5 watt Aerojet-General Nucleonics (AGN) solid homogeneous reactors to a 20 megawatt (MW) heavy water cooled and moderated tank reactor. This paper explains the ways reactors are classified by NRC to make application of a graded approach possible in regulating these facilities. Illustrative examples are presented in each area of regulation discussed. NRC applies a graded approach in all aspects of reactor regulation. The Atomic Energy Act (the Act) is the law passed by the U.S. Congress for the regulation of civilian use of nuclear technology. For research and test reactors useful in the conduct of research and development activities, the Act requires the NRC to impose only the minimum amount of regulation as the Commission finds will permit the Commission to fulfill its obligations under the Act to promote the common defense and security and to protect the health and safety of the public and will permit the conduct of widespread and diverse research and development. This requirement for minimum regulation is applied in all aspects of the regulation of research and test reactors, including licensing processes, regulatory technical requirements and inspections. NRC differentiates non-power reactors by considering two types: research reactors and test reactors (both referred to as research reactors by IAEA). The most common attribute that distinguishes between these two reactor types is thermal power level. A test reactor (also called a testing facility in the regulations) has a thermal power level greater than 10 MW. A research reactor has a thermal power level of 10 MW or less. A reactor is also designated a test reactor if it has a thermal power level greater than 1 MW and contains liquid fuel, a in-core circulating fuel test loop or a large experimental facility in the core. In general, as the risk associated with a reactor increases, the regulatory process becomes more complex with additional review required for approval of licensing actions. For example, an application for a construction permit for a testing facility is required by regulation to be referred to the Advisory Committee on Reactor Safeguards (ACRS) for a review and report. The ACRS is a statutory independent advisory committee to the Commission. A construction permit application for a research reactor is not required by regulation to be referred to the ACRS. Moreover, as the risk associated with a reactor increases, so do regulatory technical requirements. Nuclear power plants have minimum requirements for the principal design criteria which are described in the regulations. Research and test reactors are not subject to such minimum regulatory design requirements. This allows the flexibility for a wide variety of designs, experimental facilities and programs. The primary regulatory requirements for the design of research and test reactors are maintaining radiation doses to reactor staff and members of the public within acceptable limits during both normal operation and accident conditions. The NRC staff has issued documents which provide guidance on the design of research and test reactors, such as NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, which present a way an applicant can meet the regulations but alternatives are also acceptable provided they achieve the same outcomes. The research and test reactor inspection program also follows a graded approach with three classes of reactors. Class 1 reactors have a thermal power level at or above 2 MW, Class 2 reactors have a thermal power level below 2 MW and Class 3 are reactors that are permanently shut down. The 2 MW demarcation is generally based on reactor core decay heat generation and the ability to dissipate this decay heat without an active emergency core cooling system. The period of time to carry out the inspection program and the scope of inspection varies with reactor class. For example, normally the Class 1 inspection program is carried out over one year and the Class 2 inspection program over two years. Inspectors normally examine a greater number of records for a particular inspection area at a Class 1 reactor than at a Class 2 reactor. The full paper will contain greater detail on the application of the graded approach.

Speaker: Mr Alexander Adams Jr (U.S. Nuclear Regulatory Commission)

Paper IAEA_Graded_Approach_paper_A_Adams.pdf

Slides 106_Adams.pptx

11:30 PNRA Role for Ensuring Safety of Research Reactors

Operating experience of research reactors in Pakistan is spread over a span of five decades starting with the operation of Pakistan Research Reactor-1 (PARR-1) in 1965. Research reactors are being used in Pakistan for the purpose of training and production of radioisotopes. Pakistan Nuclear Regulatory Authority (PNRA) is entrusted to ensure that a high level of safety is maintained through an inclusive and robust regulatory oversight process. The objective is achieved by strict vigilance through authorization, review & assessment, regulatory inspections and enforcement processes to ensure that safety requirements are fulfilled. PNRA through an effective regulatory framework ensures and verifies that activities are performed in full compliance with the regulatory requirements. PNRA is currently regulating two research reactors i.e. Pakistan Research Reactor-1 (PARR-1), a swimming pool type reactor with a power level of 10 MW, and Pakistan Research Reactor-2 (PARR-2), a tank-in pool type Miniature Neutron Source Reactor (MNSR) of 30 KW. PNRA ensures that safety is maintained at research reactors through fulfilling safety requirements at par with international codes and standards. In the last few years, PNRA has revalidated the operating licences of research reactors on the basis of periodic safety reviews. In this paper, regulatory activities for ensuring safety and major challenges faced while regulating research reactors have been highlighted

Speaker: Ms Farhana Naseer (Pakistan Nuclear Regulatory Authority)

Paper 137_PNRA_Role_for_ensuring_Safety_of_Research_Reactor-Full_Paper_as_per_IAEA_format.docx

Slides C15_Naseer.pptx

12:10 SAFETY OF RESEARCH REACTORS IN NIGERIA

Nigeria is one of many countires that are developing infrastructure in nuclear energy. Challenges in developing its safety infrastucture are thus enormous. The Nigerian Government under the Nuclear Safety and Radiation Protection Act 19 of 1995 established the Nigerian Nuclear Regulatory Authority as the regulatory body responsible for nulear and radiological protection and also the regulation of safety infrastucture for nuclear facilities. One of Such facility is the Nigerian Research Reactor (NIRR-1). This paper seeks to look at the NNRA regulatory control regime aimed at ensuring the safety of research reactors in Nigeria. This involves a series of processes from the legal obligations, development of rgulations and guidance documents, review and assessment of documents, inspection program and enforcement actions. The challenges encountered in enforcing the control regime and possible solutions proferred are enumerated in this paper. Furthermore, it enumerates plans to ensure that all future research reactors are integrated under the regulatory control regime of the NNRA.

Speakers: Mr Julius Ejembi (Nigerian Nuclear Regulatory Authority), Mrs Moyosola Abubakar (Nigerian Nuclear Regulatory Authority)

Paper 15_Nigeria_access.doc

12:30 → 13:30 Lunch Break

13:30 → 15:00 Utilization and Application

Convener: Mr Junaid Razvi (TRIGA Group)

13:30 Lessons Learned on Strategic Planning for Enhanced Utilization of Low Power Research Reactors

The International Centre for Environmental and Nuclear Sciences (ICENS) was set up to be an institution with a formidable range of applications that promised to impact on Caribbean expertise and its potential to advance the application of science and technology to the sustainable development of the member countries of the University of the West Indies. Before the reactor was commissioned a small, but influential group, was established called the committee for the peaceful uses of atomic energy to chart the way forward for the facility. The SLOWPOKE-2 reactor, JM-1, at ICENS is a pool type, light-water-moderated and cooled by natural circulation, with a nominal power of 20 kW (thermal). The reactor achieved first criticality in March 1984 with HEU fuel and has operated on a regular weekly schedule since then. The main use of the reactor has been neutron activation analysis, presently there are four inner (reflector) irradiation sites, one outer (external to the reflector) irradiation site and one in-pool irradiation rig. Refuelling with LEU is scheduled to take place in the last quarter of 2015. Fortunately, the dimensions of the LEU fuel are almost identical to the HEU fuel, thus allowing the reuse of all the ancillary systems. The facility is housed at the University of the West Indies in Jamaica and is one of several specialized research units on the campus. The ICENS, which is a partnership between the University and the Government of Jamaica, has a scientific agenda which emphasises integrated research programmes based on the chemistry of the total environment and its effects on the biosphere. The analysis outputs are used to develop powerful interactive geochemical databases which combined with Geographical information systems, provides a valuable resource critical in the development of evidence based policy formulation. It makes use of inter-institutional (University/Government Ministries) and international collaborations to ensure that the research activities are relevant to all stakeholders; in particular, the transfer of knowledge between academia and government with a major objective being the development of the human and economic resources of the country. These collaborations include M.Phil. and Ph.D. students from the Departments of Basic Medical Sciences, Botany, Chemistry, and Geology; MSc. Students from Forensic Chemistry, teaching undergraduate chemistry students and providing courses in Radiation Physics, Radiation Biology and Radiation protection for BSc. in Diagnostic Imaging. Although the majority of the funding for the institution is provided by the University and the Government, it also provides services for the private sector. The commissioning of the reactor necessitated the installation of, in the Jamaican context, unique pieces of equipment; these include gamma spectroscopy systems, radiation monitors, thermoluminescence dosimetry systems, liquid nitrogen plant as well as other complimentary analytical techniques. ICENS provides the only dosimetry service in the English speaking Caribbean for approximately 1500 persons; it also provides radiation leak testing services for our industrial radiographers and mining companies and liquid nitrogen to several dermatologist and other laboratories. These activities present unique opportunities that could be replicated in similar facilities, there is very little competition in the market place, and the activities are relatively simple, thus guaranteeing much needed additional income. This paper reports on the strategies and policies adopted by ICENS over the past 31 years for academic, governmental and private sector partnerships. In addition, the utilization of the SLOWPOKE-2 reactor for various research programmes along with detailed descriptions of the experimental facilities is also reported.

Speaker: Mr Charles.N. Grant (The International Centre for Environment and Nuclear Science)

Paper 188_Lessons_Learned_on_Strategic_Planning_for_Enhanced_Utilization_final.docx

14:00 Improved performance of neutron activation analysis laboratories by feedback meetings following interlaboratory comparison rounds.

Over the years, the IAEA has stimulated the orientation of neutron activation analysis (NAA) laboratories worldwide on fields of application in which a large number of samples may exist for analysis. Whereas the markets for service by NAA laboratories may have been identified, demonstration of valid analytical data and organizational quality of the work process are preconditions for consolidating and expanding the stakeholder community. Eventually, laboratories and/or stakeholders may prefer that the facility’s management system is accredited for compliance with the International Standard ISO/IEC17025:2005. One of the requirements in the process towards such accreditation is that the laboratory provides evidence of the validity of its measurement results by participation in proficiency testing schemes by interlaboratory comparison. Participation in interlaboratory comparison study may reveal that some results are not satisfactory. Laboratories are then facing the problem of finding the source of such non-conformity and applying corrective actions. Obviously, providers of intercomparison rounds cannot provide such a laboratory and technique-specific after-care. The IAEA has therefore implemented a new mechanism for supporting the NAA laboratories in their Member States in demonstrating their analytical performance by assisting them in identifying unanticipated sources of error, to assess with them approaches for elimination thereof and to design with them a path for growing towards sustainable performance at the analytical state of the practice. This was accomplished by an evaluation and feedback meeting following the participation in proficiency testing by interlaboratory comparison. Laboratories under the IAEA Technical Cooperation (TC) projects RAF4022, RAS1018, RER4032/RER1007 and RLA0037 participated, facilitated by the IAEA, between 2010 and 2013 in consecutive proficiency testing schemes by interlaboratory comparison of the Wageningen Evaluating Programs for Analytical Laboratories (WEPAL) to assess their analytical performances. WEPAL, a provider of proficiency testing schemes, is accredited by the Dutch Council for Accreditation for compliance with the International Standard ISO17043:2010. The results have been analysed by IAEA experts providing first indications for potential sources of error, and further discussed by experts and participants in feedback meetings. This IAEA’s initiative to facilitate laboratories participating in proficiency testing schemes complemented by the new approach of feedback meetings resulted in a significant increase in the analytical and associated organizational performance of most participating laboratories. Several other laboratories demonstrated consolidation of their already satisfactory performance. The increase in performance was achieved by an increase in awareness on potential sources of error, technical and/or organizational, and related approaches of quality control and quality assurance to be implemented. Recently the initiative has been extended to all regions resulting in 35 NAA laboratories world-wide who are signed for 2015 round tests, with the follow up workshop scheduled in August 2015.

Speaker: Dr Peter Bode (Delft University of Technology)

Paper Paper_Bode-Ridikas_CN231_18.pdf

14:20 Overview of NAA method applied at Es-Salam research reactor

At Es-Salam research reactor, the instrumental neutron activation analysis was used since 1993 as the first technique in our laboratory. In spite of the simplicity of the concept of this technique, the capability of analysis depends of the determination of elements in reference standard. Due to the limitations of INAA method, we have conducted our objective towards the well-known advantages of the k0-NAA method. A project for introduction of this method was conceived in 2002 and implemented two years later. The k0-NAA procedure established in our laboratory has been regarded as a reliable standardization method of NAA and as available for practical applications. Examples of such samples, within a selected group of disciplines are milk, milk formulae, salt, plants and seeds (nutrition), hair and blood of human (clinical investigation), cigarette tobacco (environmental and health related fields) and iron ores and clays (exploration and mining). In addition, the development of other analytical techniques based on the neutron activation is also achievement during 2005 such as the cyclic delayed neutron counting technique and the application of RNAA to the proportioning of iodine in food salt. The NAA laboratory work to make an effort to connect the unique features of NAA activities in a strategic way for the national goals by its accreditation. Our experience on the inter-laboratory exercises undertaking in the frame work of RAF4020, RAF4022 and RAF1005 projects for the evaluation of the analytical laboratory competency was also discussed.

Speaker: Mrs Lylia Hamidatou (Nuclear Research Centre of Birine)

14:40 Advances in utilisation of the JSI TRIGA Mark II reactor

The TRIGA Mark II reactor at the Jozef Stefan Institute (JSI) is regularly used for Neutron activation analysis (NAA) in both instrumental (INAA) and radiochemical (RNAA) modes, radiation hardness studies, training of NPP staff and education of university students, and for benchmark experiments for validation of computer codes and nuclear data. In the last 5 years the scientific level as well as the intensity of these activities has increased significantly. The paper describes the latest advances in the utilisation of the JSI TRIGA Mark II reactor. The reactor has been extensively used for irradiation of various components to study radiation hardness for the ATLAS detector in the European Organisation for Nuclear research (CERN) and other accelerators under the AIDA framework, for ITER and for some companies producing radiation tolerant products. Due to very good characterization of the irradiation facilities the reactor has become a reference centre for neutron irradiation of detectors developed for the ATLAS experiment. Due to large flexibility and well defined irradiation condition the irradiation facility has recently been developed, which allows irradiation in a well thermalized (99.8 % of total neutron flux is thermal) neutron flux. In collaboration with the CEA/DAM Ile de France the facility is used for performing neutron irradiation of materials used for control of non-proliferation in nuclear safeguards by using the FT-TIMS method. In 2010 a collaboration with CEA Cadarache Instrumentation, Sensors and Dosimetry Laboratory was established leading which lead to several experiments of benchmark quality; fission rate profile measurements, neutron dosimetry and spectra measurements, beta-effective measurements, gamma profile measurements, self-powered neutron and gamma measurements. These experiments were used for validation of computer codes and nuclear data and lead to better characterisation of irradiation conditions. In addition several CEA-developed detectors and data acquisition systems were tested within the collaboration. In the field of education and training the reactor is used in regular laboratory exercises for graduate and post graduate students of physics and nuclear engineering at the Faculty of Mathematics and Physics, Ljubljana University and for training of operators at the JSI Nuclear training Centre. Since 2010 new advanced practical exercises for students and course attendees were designed and implemented; void reactivity coefficient measurements, the measurement of water activation, in-core flux mapping system, pulse mode operation, etc. The reactor team takes active part in the Eastern Europe Research Reactor Initiative [EERRI] and the Mediterranean Research Reactor Network [MRRN], supported by the International Atomic Energy Agency (IAEA). The JSI TRIGA Mark II is a good example of a relatively small research reactor having a rather low neutron flux that can support a variety of high-level research and developments activities as well as education and training at a very high level. In the paper the abovementioned activities are presented in more details.

Speaker: Dr Luka Snoj (Jozef Stefan Institute)

Paper 2015_11_IAEA_RR_041-paper_recent_advances_in_utilisation_of_JSI_TRIGA_reactor_rev-anne_laure.pdf

15:00 → 16:00 Break: Poster Session and Coffee Break

16:00 → 18:00 Operation and Maintenance

Convener: Mr John Arnold (Canadian Nuclear Laboratories)

16:00 THERMAL POWER CALIBRATION AND NEUTRON FLUX MEASUREMENT OF THE NUCLEAR RESEARCH REACTOR IAN-R1 2015

ABSTRACT The IAN-R1 TRIGA (Training Research Isotopes General Atomics) Nuclear Research Reactor is operated by the Colombian Geological Survey (CGS) located in Bogota, this is a pool type reactor and operates at 30 kW. During the year 2012 the reactor underwent a major upgrade on its instrumentation and control systems, task that couldn’t have been carried out without the help of the “Instituto Nacional de Investigaciones Nucleares” (ININ) from Mexico. Following this upgrade a series of flux measurements and power calibration were performed, experiments which will be presented on this paper. Neutron Flux mapping is essential to the Reactor in order to offer accurate irradiation services. The calorimetric method was used to carry out a Power Calibration with the Reactor operating at 20kW. Five (RTD) thermocouples were positioned axially above the core, and insulation was used to cover the reactor’s pool and minimize heat transfer to the surroundings. The Nuclear Reactor was operated for 4 hours and 45 minutes for the experiment, and temperature data was taken every 15 minutes. A linear dependency of temperature vs time was observed, and the reactor’s power was obtained from the slope of the curve. The reactor’s core consists of a square grid in which fuel rods are located. These are low- enriched uranium TRIGA fuel rods (up to 20% U235). Four fuel rods make up a cluster, and each cluster is positioned on a different position on the grid making up the core. Some clusters contain three fuel rods and control rods, or irradiation channels depending on the position in the core.There are three control rods and three pneumatic irradiation channels available within the core. The Neutron’s thermal flux was measured axially with an energized vanadium detector along the positions F1, F2, F3, F4, F5, F6 and at the midpoint on row G between positions 3 and 4. Small gold and aluminum particles (0.1% Au, 99.9% Al) were irradiated and then analyzed by Neutron Activation Analysis. We can conclude that proper calibration of reactor power and flow measurement of thermal neutrons has been demonstrated in the good performance obtained by the Laboratory of Neutron Activation Analysis (CGS) who radiate samples in this reactor during the first round of proficiency testing Wageningen Evaluating Programs for Analytical Laboratories (WEPAL) received in 2015.

Speaker: Mr Jaime Sandoval (Servicio Geológico Colombiano)

Paper Power_Calibration_Paper_Nov_2015.pdf

Slides 134_Presentation_Power_Calibration_Format.ppt

16:20 Reactor PIK: the upgraded core

The PIK Reactor is currently under preparation for its power start-up. Its core allows the reactor to achieve the designed power of 100MW. At the same time, designed cycle duration of two weeks is not long enough for the present day experiments. Another reason for the core upgrade is the lack of volume for advanced irradiation of surveillance specimens. Operating reactor cycle of around 30 days is to be obtained.

Speaker: Dr Kir Konoplev (Petersburg Nuclear Physics Institute)

Paper Report_для_PDF_ver.2.pdf

Slides D11_Konoplev.pptx

16:40 Advances in the Development and Testing of Micro-Pocket Fission Detectors

Miniaturized sensors capable of real-time neutron flux measurements are needed for in-core deployment in research nuclear reactors. Prototype Micro-Pocket Fission Detectors (MPFDs) have previously been tested in neutron beams up to 10^8 n cm^-2 s^-1 [1] and in-core up to an estimated neutron flux of 8×10^12 n cm^-2 s^-1 [2]. Although previous studies confirmed MPFDs can be used to track reactor power, improvements were needed regarding detector dead-time and neutron-reactive material deposition methods. MPFDs hold several advantages over conventional neutron flux measurement techniques. Typical ionization and fission chambers are large and are often used externally to monitor neutrons which have escaped the reactor core [3]. Many fission chambers are composed of highly-enriched U-235 [4], and those designed for high-flux applications are typically operated in current-mode due to the high interaction rate [5]. Such devices are impractical for use in critical mock-ups, high performance material test reactors (MTRs), and transient test reactors because of their fragile construction and large flux perturbation when installed in-core or near-core. MPFDs, however, are constructed of radiation-resistant materials and do not significantly perturb the local neutron flux. MPFDs also provide continuous, real-time pulse-mode measurement capability for extended in-core operation. Significant progress has been made to advance the development of MPFDs. A controlled electro-deposition process for neutron-conversion materials has been developed. Non-destructive measurement techniques have been utilized to measure the mass of neutron-conversion material deposited onto electrodes much smaller than 1.0 mm^2. Finally, optimization of neutron-conversion material compositions extends stable device operation for high-fluence applications [6]. MPFDs presently utilize small depositions (< 1 μg) of natural uranium yielding a very low interaction rates (~10^-8 fissions per neutron), enhancing operation in high-neutron-flux environments. Advanced MPFDs have been built and deployed in the Kansas State University TRIGA Mk II research reactor (neutron flux up to 10^12 n cm^-2 s^-1). Detector parameters such as ionization chamber volume, neutron-reactive material mass, and electrode size can be varied based on application. The most recent development in MPFD research has produced small neutron detectors capable of pulse-mode operation in the high-neutron and high-gamma-ray flux of the reactor core [7]. These advanced MPFDs are inherently gamma-ray insensitive due to their small size and have operated in pulse-mode up to a reactor power of 200kW (~10^12 n cm^-2 s^-1). Research is ongoing to develop MPFD arrays capable of simultaneous, real-time measurement of the neutron flux in multiple locations throughout a reactor core. Detector chambers which are sensitive only to fast neutrons are also being developed. Electronics packages are also in development which will standardize device readout and eliminate the need for numerous NIM components. Ultimately, the compact, accurate, high-fluence neutron detectors will be deployable for critical mock-ups of existing and advanced nuclear reactors designs, high-performance materials test reactors, and transient test reactors. References [1] M. F. Ohmes, D.S. McGregor, J. K. Shultis, P. M. Whaley, A. S. M. S. Ahmed, C. C. Bolinger, T. C. Pinset, "Development of Micro-Pocket Fission Detectors (MPFD) for near-core and in-core neutron flux monitoring," Hard X-Ray and Gamma-Ray Detector Physics, pp. 234-242, 2004. [2] M. F. Ohmes, A. S. M. S. Ahmed, R.E. Ortiz, J. K. Shultis, D. S. McGregor, "Micro-Pocket Fission Detector (MPFD) performance characteristics," in IEEE Nuclear Science Symposium, 2006. [3] N. Tsoulfanidis, "Measurement and Detection of Radiation," Taylor and Francis, Washington, 1995. [4] Ch. Blandin, S. Breaud, L. Vermeeren, M. Weber, "Development of New Sub-Miniature Fission Chambers: Modeling and Experimental Tests," Progress in Nuclear Energy, vol. 43, no. 1-4, pp. 349-355, 2003. [5] S. P. Chabod, A. Letourneau, P. Gourdon, C. Laye, "Improvements in the Modeling of Sub-Miniature Fission Chambers Operated in Current Mode," IEEE Transactions on Nuclear Science, vol. 57, no. 5, pp. 2702 - 2707, 2010. [6] M. A. Reichenberger, P. B. Ugorowski, J. A. Roberts, D. S. McGregor, "First-Order Numerical Optimization of Fission-Chamber Coatings Using Natural Uranium and Thorium," in IEEE Nuclear Science Symposium, Seattle, WA, 2014. [7] T. Unruh, J. Daw, K. Davis, D. Knudson, J. Rempe, M. Reichenberger, P. Ugorowski, D. McGregor, J. Villard, "Enhanced Micro Pocket Fission Detector Evaluations," in NPIC HMIT, Charlotte, NC, Feb. 23-26, 2015.

Speaker: Mr Michael Reichenberger (Kansas State University)

Paper ID_103.pdf

17:00 OPERATIONAL APPROCH ON AGEING MANAGEMENT AT THE GHANA RESEARCH REACTOR -1 FACILITY

Attached is a copy of the synopsis of my presentation

Speaker: Mr Henry Obeng (Ghana Atomic Energy Commission)

Paper VEINNA_15-OPERATONAL_APPROCH_-Document.pdf

Slides 5_PPT-Operational_Approach.pptx

17:20 The “Rosatom” activity in the field of research reactors

The Russian Research Reactors especially high power reactors concentrated mainly in institutes of “Rosatom”. The utilization of them is very high and load factors during many years are not less than 0.6. The reactors have a different power level, neutron flux and spectrum and it get the broad opportunities for scientific and applied works. The main areas of reactor utilization are material testing and radioisotope production. It is important that at the institutes that have research reactors there are also good equipped hot cells and it allows to carry out the broad post-irradiation examinations. The safety is a high priority in the operation of research reactors of “Rosatom”. During the analyses of the accident at the Fukushima Daiichi NPP to the Russian research reactors the role of the “Rosatom” as the state organization was clear expressed. This role was very important in the initiating activity to carry out stress-tests and financial support of measures for safety increasing. The deep safety analyses of these reactors were made after the accident at the Fukushima Daiichi NPP and necessary measures were implemented for enhancing the reactor safety. The “Rosatom” develop the activity in the construction of new reactors (MBIR project) and refurbishment of operating facilities (critical assemblies BFS in IPPE). Last years “Rosatom” continue to develop the international cooperation in the development of research reactors and Centers of Nuclear Research. Such proposals were made for several countries and projects are in progress.

Speaker: Mr Nikolay Arkhangelsky (Rosatom)

17:40 Dedicated facility for mass-producing of doped Silicon

Nowadays, the use of electric vehicles such as hybrid cars and electric trains is increasing rapidly. Each such vehicle requires a considerable amount of semiconductor devices, suggesting an ever-growing demand for such devices in the near future. Thus, the mass production of the semiconductor material is becoming a very important issue for many manufacturers. There are several methods for mass production, but Neutron Transmutation Doping of Silicon (NTD-Si) is the most promising method for the production of silicon based semiconductor. Rapid increase in demand of hybrid cars is expected due to the high cost of fuel and environmental concerns. A recent survey made by Korean Atomic Energy Research Institute (KAERI) indicated that 50 million hybrid electric vehicles would be produced, and almost 1000 tons of NTD-Si would be needed in 2030. At present, the worldwide capacity of the NTD-Si facilities is estimated to be 150-180 tons per annum. This capacity cannot be increased drastically because most research reactors with NTD facilities were constructed many years ago and they have little potential for expansion; in addition, number of research reactors to be constructed in the future is few. Another important issue is that the present research reactors are not NTD-dedicated facilities, and thus a stable and adequate supply is not expected. If the hybrid electric vehicles increase according to the current forecast, the existing research reactors cannot supply future demand. As such, a new doping facility with a large irradiation capacity for NTD-Si may need to be constructed to ensure an adequate supply of doped silicon. Two design concepts were proposed. The neutronic and thermal hydraulic analyses were performed to obtain the optimum core composition, operating period, necessary condition for uniform doping and the reactor production rate. First one was a design concept of a small reactor for large-diameter NTD-Si using full-length conventional PWR fuel assemblies. The idea was to use commercially available conventional PWR fuel (fabricated for power reactors) directly without any modification. Estimated production rate varies between 111 tons/year and 140 tons/year for 50 Ωcm target resistivity depending on the control rod positions. Second one was a design concept of a small reactor for large-diameter NTD-Si using short PWR fuel assemblies. Estimated production rate varies between 48 tons/year and 70 tons/year for 50 Ωcm. The first concept has a very good performance, for instance, a long operating period (about 18 years) and a large production rate, but the size of the core could become big which could lead a large construction cost (initial investment). On the other hand, the reactor concept using short PWR fuel assemblies might require a less construction cost due to the smaller core, and the production rate was not that bad, but the operating period was shorter (about 3 years) than first concept. This concept with short assemblies may be attractive if the shorter PWR assemblies can be fabricated without major difficulties in existing fuel manufacturers, needless to say that they are technically capable to produce such assemblies. Another important point is that fuel supply for these concepts is more reliable and fuel management is easier because PWR fuel fabrication and processing facilities well established compared to the research reactor.

Speaker: Dr MUNKHBAT Byambajav (Nuclear Research Center, National University of Mongolia)

Friday, 20 November

09:00 → 10:30 Utilization and Application

Convener: Mr Jose Augusto Perrotta (CNEN)

09:00 Issues and Challenges in Research Reactors based Radioisotope Production

Research reactors play a key role in the production of radioisotopes for various applications in nuclear medicine (Mo-99/Tc-99m, I-131, Xe-133, Lu-177, Ir-192, Ho-166, P-32, Y-90, Sm-153, Re-186, W-188/Re-188, Sr-89, Sn-117m, Ac-227/Ra-223, …), industry (Ir-192, Se-75, Hg-203, …) and research. While the number of accelerators – mainly cyclotrons – is increasing specially for the production of medical radioisotopes, the supply of reactor-produced medical radioisotopes relies on a limited number of research reactors. This is the case for the production of Mo-99, a very crucial radioisotope as it decays into Tc-99m which is used in 80% of the 30 million radiodiagnostic nuclear medicine procedures carried out worldwide annually. The current situation is a major concern in the Mo-99/Tc-99m supply chain, especially after the decisions taken recently to shut down definitively the OSIRIS reactor (France) in December 2015 and to cease routine Mo-99 production at the NRU reactor (Canada) from November 2016. Current research reactors are ageing, expensive to replace and due to safety and financial issues, it is a continuing source of public and political debate. Their availability with appropriate neutron fluxes, significant operating time and economic viability are important issues to ensure a secure and reliable supply of radioisotopes in future. Successful conversion of High Enriched Uranium (HEU) into Low Enriched Uranium (LEU) for reactor fuel and targets for Mo-99/Tc-99m production are also important challenges in the coming years in the frame of the National Nuclear Security Administration Global Threat Reduction Initiative. In this context, the Belgian Nuclear Research Centre (SCK•CEN) made a strategic decision to refurbish the BR2 reactor, which is considered as a major facility worldwide for the routine supply of Mo-99/Tc-99m, I-131, Xe-133, Lu-177, Ir-192, Re-186, Sm-153, Er-169, Y-90, P-32, W-188/Re-188, I-125, Sr-89, Sn-117m, … and for the development of new medical radioisotopes as Ac-227/Ra-223, ... The BR2 reactor is currently in temporary shutdown for a scheduled period of 16 months from February 2015 until June 2016 to replace mainly its beryllium matrix. The refurbishment will allow a safe and reliable operation of the reactor for another period of at least 10 years with an upgraded annual operating regime of up to 8 cycles, i.e. up to 180-200 operating days per year, subject to the economics.

Speaker: Dr Bernard PONSARD (SCK.CEN - BR2 REACTOR)

Paper 183_PONSARD_PAPER_Issues_and_Challenges_in_Research_Reactors_based_Radioisotope_Production.docx

09:30 Production of radionuclides for medical use at WWR-c reactor, radiopharmaceutical production under GMP standards

At WWR-c reactor of Karpov Institute of Physical Chemistry, Obninsk, the following radionuclides for medical use are produced: 99Мо, 125I, 131I, 135Xe, 153Sm, 59Fe. Also there is a production of other radionuclides at WWR-c reactor, including radionuclides for scientific research and development or for non-repeat orders. However, there is no great demand for the other medical radionuclides to be produced at the reactor. At present the most crucial task is to increase the production of one of the most high-demand radionuclides 99Мо for diagnosing of oncological diseases. Due to failure of starting two 99Мо reactor-generators in Canada, the lack of this isotope has appeared [1]. This fact significantly raised the chances of Karpov Institute to stand with Russian 99Мо on the world market. That is why the task to increase the production of the above mentioned radionuclides and generators on their base with simultaneous properties improvement is an important economic goal. Experience of 99Мо production at WWR-c reactor. In 1983 Karpov Institute accepted a challenge to set up 99Мо production. One channel at reactor core periphery was selected for that purpose. After the force-cooled channel was designed and target design in 1985 the first batch of 99Мо was produced. The volume of 99Мо production was about 20 Ci per week [2]. In 2002 the target was modified and the number of targets in the channel was raised up to 4 pieces. Also the procedure of target processing was optimized and the coefficient of 99Мо radiochemical release was increased from 60 up to 85 %. Thus, the channel capacity for 99Мо production grew up to 30 Ci per week. This production capacity was enough for the full cover of Russian hospitals in existing 99Мо demand. At the end of 2008, one additional channel was installed in cell 8-1 of WWR-c reactor core that allowed to double the 99Мо production starting from 2009. Thus, starting from 2009 Karpov Institute could produce both 30 Ci per week (commercial) required for local hospitals and up to 50 Ci per week of 99Мо for export shipments. Since 2012 two more channels have been installed. This allowed to double the existed production of 99Мо. In 2015 the current volume of 99Мо production is 200 Ci per week (commercial). The potentiality of 99Мо production at WWR-c reactor went up to 400 Ci per week, what is to be realized after the advancement of the radiochemical process. In 2016 it is intended to perform the production modernization (irradiation, radiochemical release). Production line for 99mTc-generators loading. As of today Karpov Institute weekly produces 100 – 130 generators with different nominal activities. In order to decrease the possibility of human mistakes during the process of generators loading and to increase their quality and also according to the new GMP requirements for drug manufacturing Karpov Institute decided to optimize Karpov generators production by changing manual loading to automatized one. The new production line shall correspond to the following main requirements: • Automatic operation with minimal operator intervention; • Production volume should equal to 200 generators per week; • Production equipment shall be installed in rooms which correspond to GMP requirements. During 2007 – 2009, one of Karpov Institute buildings was refurnished to construct “clean” rooms with production line for automatized loading of molybdenum-technetium generators under GMP (Good Manufacturing Practice) standards. In 2013, the installation of the major equipment was finished and pre-commissioning activities were performed [3]. At the beginning of 2015 the cold startup of the line was performed. Karpov Institute together with the engineering company trains the Institute’s specialists in the field of loading of molybdenum-technetium generators. The estimated period of commercial commissioning of this facility is 2015 year. As of today the specialists undergo training on working in compliance with GMP standards. At present the development of the project on the conversion of additional buildings of the Institute to manufacture other radiopharmaceuticals under GMP standards is in progress at Karpov Institute. REFERENCES 1. N. Ramamoorthy “Overview of 99Mo Production Crisis and Options for Improving Supplies”// TWG-RR Meeting - 2010. 2. V.V. Pozdeev Izotopy eto slozhno, no nuzhno [Isotopes – it is hard, but it is necessary] // AtomInfo.ru – 2010. 3. O.Yu. Kochnov Obninskie generator technetsiya [Obninsk technetium generators] // AtomInfo.ru – 2013.

Speaker: Mr Vladimir Gremiachkin (Karpov Institute of Physical Chemistry)

Paper ID_87.docx

Slides Presentation_IAEA-CN-231-87_Gremiachkin_V..ppt

09:50 Current Status of Operation and Utilization of Dalat Nuclear Research Reactor and Strategic Plan in the Next Decade

Dalat Nuclear Research Reactor (DNRR) with the nominal power of 500 kW is today the unique one in Vietnam. The reactor was reconstructed and upgraded from TRGA Mark II research reactor, and restarted operation on March 20, 1984. Under the framework of the program on Russian Research Reactor Fuel Return (RRRFR) and the program on Reduced Enrichment for Research and Test Reactor (RERTR), the project for full core conversion of the DNRR from high-enriched uranium fuel (HEU) to low-enriched uranium fuel (LEU) was implemented during years 2008 - 2012. Since January 2012 the reactor has been operated with a working core configuration consisting of 92 VVR-M2 LEU fuel assemblies of 19.75% enrichment. Up to mid 2015, the reactor has been operated with the total of about 40,250 hrs of safety and effective exploitation. During the last 31 years of operation, the DNRR was efficiently utilized for: (1) Producing about 400 Ci per year of radioisotopes including I-131, P-32, Tc-99m generator, Cr-51, Sm-153 for medical use; (2) Developing a combination of nuclear analysis techniques (INAA, RNAA, PGNAA) and physic-chemical methods for quantitative analysis of about 70 elements and constituents in various samples of geology, crude oil, agriculture, biology, environment; (3) Carrying out experiments on the reactor horizontal beam tubes for nuclear data measurement and nuclear structure study; and (4) Contributing on nuclear education and training programs for human resource development in the country. In the next ten years, the DNRR will be considered to implement the operation regime of 150 hrs/cycle in order to increase the quantity of radioisotope production as well as the number of irradiation samples for NAA. The use of the reactor for nuclear education and training in order to support the human resource development for the national nuclear power program will be also paid much attention. Besides, in preparation for the effective utilization of a new research reactor in future, the broadening of researches on production of various radioisotopes (such as Mo-99, Y-90, Ho-166, Lu-177 and Ir-192 ), neutron radiography and gemstone coloration will also be planned to carry out at the DNRR. As the high power research reactor puts into operation between 2023-2025, the DNRR will shift its utilization purpose and mainly use for NAA, basic researches, and education and training. This paper presents the current status of operation and utilization of the DNRR. In addition, the strategic plan for the reactor in the next decade is also mentioned in the paper. Keywords: DNRR, HEU, LEU, RRRFR, RERTR, VVR-M2, NAA, INAA, RNAA, PGNAA, BNCT.

Speaker: Mr BA VIEN LUONG (Nuclear Research Institute)

Paper Current_Status_of_Operation_and_Utilization_of_DNRR_and_Strategic_Plan_Vietnam.pdf

Slides Current_status_of_Operation_&_Utilization_of_DNRR_and_Strategic_Plan_Vietnam_2015.ppt

10:10 Role of BAEC Research Reactor in the Development of Nuclear Science and Power Programs in Bangladesh

The Bangladesh Atomic Energy Commission (BAEC) has been operating the TRIGA research reactor since September 1986. The BAEC TRIGA Research Reactor (BTRR) is the only nuclear reactor of the country. The reactor has been used for manpower training, education, radioisotope production and various R&D activities in the field of neutron activation analysis, neutron radiography, neutron scattering and experimental research on reactor safety. The BTRR is a light water cooled, graphite reflected reactor, designed for maximum steady state thermal power level of 3 MW and for pulsing operation with maximum pulse power of 852 MW. Center for Research Reactor (CRR) of BAEC is responsible for operation and maintenance of the reactor as well as human resources development in the area of nuclear power program. Bangladesh government has strong commitment to implement nuclear power programs (NPP) in the country. An agreement was signed on 2 November 2011 between the Government of the Russian Federation and the Government of the Peoples Republic of Bangladesh for the construction of a NPP (2000 MW) on the territory of Bangladesh. On 2 October, 2013 the Honorable Prime Minister, the Peoples Republic of Bangladesh has laid the foundation stone of Rooppur Nuclear Power Project (1st phase) implementation activities. The Rooppur site is north-west region of Bangladesh and this would be the first nuclear power plant in the country. The knowledge and experience gain from operating the research reactor, directly or indirectly can support the development and implementation of nuclear power programs. The experience from managing nuclear material at research reactors promotes a better understanding of the infrastructure and issues that need to be addressed in the field of nuclear power programs. Nuclear reactor technology related training and education program has been extended to provide necessary supports to the students undertaking nuclear engineering courses in various public universities of the country. Conducted different practical experiments at the CRR on the following topics: Measurement of radiological safety parameters; reactor operation and maintenance; Thermal hydraulics related experiment; measurement and study of reactor physics and nuclear safety parameters; calibration of different reactor safety related equipment; study on digital I&C systems and reactor safety related equipment, etc. The reactor facility along with the associated laboratories has been used successfully for carrying out routinely thesis works in the field of nuclear science and technology of B.Sc./MSc./MPhil/Ph.D. students from different public universities of the country. The reactor facility has been used for training and retraining programs of the reactor operating personnel (including foreigners) to the level of Senior Reactor Operator (SRO) and Reactor Operator (RO). The facility also arranges several practical experiments on nuclear safety related parameters for the participants of different training courses such as Basic Nuclear Orientation Course (BNOC) as foundation training course of the newly appointed scientists and engineers, Fundamental course on nuclear power plants, Follow-up Training Course on reactor engineering, etc. For enhancement of utilization of the reactor, a strategic plan has been developed for BTRR. The plan has identified facility’s strengths, achievements, weaknesses, opportunities and threats, strategic issues and prepares a time bound action plan for achieving the goals. BAEC with its limited resources is always trying hard to strengthen the safeguards and physical protection programs around its research reactor and associated facilities. The BAEC plays a leading role in the planning, implementation, and evaluation of the nuclear safeguards and security activities in different nuclear and radiological facilities (e.g., Research Reactor, Central Waste Processing and Storage Facility (CWPSF), 60Co source of the Institute of Food and Radiation Biology of AERE and Radiation Oncology Centers in the country. There are 19 issues consider in infrastructure building for NPP. Most of these issues are common for NPP and research reactor. BAEC has been operating a research reactor for about 3 decades. The major potential areas of research reactor contribution to the building for NPP are: Nuclear safety, regulatory activities, safeguards, radiation protection, human resource development environmental protection, emergency planning, security and physical protection, nuclear fuel handling and storage, radioactive waste, etc. Most of these issues are handling by the BAEC and regulatory personnel successfully. The supporting infrastructures, experience and expertise by the existing research reactor would be helpful for taking knowledgeable decision regarding NPP. Research, education and human resource development programs enhance significantly with view to implement the NPP. BTRR is playing an important role for human resources and infrastructure development for nuclear science and nuclear power programs in the country.

Speaker: Dr MD. ABDUS SALAM (Research Reactor)

Paper Full_Paper_on_Reactor_Utilization_Bangladesh.docx

10:30 → 11:00 Break: Poster Session and Coffee Break

11:00 → 12:30 Common Management Considerations

Convener: Ms Mirela Gavrilas (U.S. Nuclear Regulatory Commission)

11:00 Interface between Safety and Security of Research Reactors

The operation of a reliable and well utilised research reactor facility is dependent on many factors, and the basis of all the factors are the dual platforms of safety and security. Good design, well planned and executed operations and maintenance, healthy and strong cultures, effective leadership and management, appropriately trained staff and many other factors are all enabled and based on strong and effective safety and security. The overall objective for nuclear safety and nuclear security is the same, being: protection of people and the environment from the harmful effects of radiation. Safety and security share the common goal of the elimination of risk. The mechanisms for achieving the common objective are different from the perspectives of safety and security. This partly stems from the examination of the risks associated with safety and those from security, which demonstrates that those risks emanate from different sources. Safety is about avoiding, protecting and mitigating accidents, while security is about protecting against, and mitigating, threats. For safety the overriding strategy to combat risk is “defence-in-depth” against accidents which, to be effective, needs to be embedded in design and operations, while the complementary approach for security is “detection, delay and response” to threats. Understanding the distinctions of these sources of risk can help operators, regulators and users become more effective in their domains so that research reactors can be utilised to their full potential. Being able to delineate the interface between safety and security is an important facet of managing a research reactor facility effectively. Both nuclear security and nuclear safety have distinctive cultures, which share many common characteristics, have elements which are different, and to be most effective are required to coexist. The paper will examine the elements of the coexisting cultures so that the interface of safety and security can be detailed. Practical examples will also be used from our experience with the OPAL research reactor. The OPAL research reactor located in Sydney, Australia is operated on behalf of the Australian Government by the Australian Nuclear Science and Technology Organisation (ANSTO). OPAL is one of the world’s more modern high powered research reactors, and the process for licencing, and operation considered safety and security matters according to best international practice. The mission for OPAL is multi-purpose in nature, with the main components being scientific and engineering research using neutron beam scattering and imaging techniques and irradiation facilities which are part of the supply chain for various radio-pharmaceuticals, research based radio-isotopes, neutron-transmutation doped silicon, and neutron activation analysis techniques. In order to firstly obtain, and secondly maintain a license to operate, ANSTO was required to demonstrate the safety and security of the reactor via processes which involved submissions, reviews and approvals by two regulatory bodies – the Australian Radiation Protection and Nuclear Safety Agency (ARPANSA) and the Australian Safeguards and Non-Proliferation Office (ASNO). The interplay between the two regulatory bodies and the operating organisation was an important factor in determining outcomes in the security area for which there was no available international precedent. Safety and security issues were considered at various levels of detail and depth in the applications for siting, construction, and operation. Once an authority to hot commission was obtained through the issue of the OPAL operating licence, requirements were made through licence conditions for periodic reviews of safety and security. The first periodic reviews for safety and security have now been undertaken and have received the necessary approvals from ARPANSA and ASNO. In March 2015 the OPAL licence conditions were modified to explicitly include development of a program supporting continuous improvement in safety culture, analysis of the interdependencies between safety and security of nuclear and other radioactive materials, and development of an overarching protective security system. Since the operating licence for OPAL was granted, there have been other significant reviews undertaken in the safety and security areas following specific external and internal events, for example following the Fukushima accident. The experiences from these reviews will also be covered in the final paper and presentation.

Speaker: Dr Greg Storr (Australian Nuclear Science and Technology Organisation)

11:30 Graded approach applications in nuclear research reactors

A graded approach is applicable in all stages of the lifetime of a research reactor (site selection, site evaluation, design, construction, commissioning, operation and decommissioning). The IAEA Safety guide no. (SSG-22) presented recommendations on the graded approach to application of the safety requirements for research reactors. In these applications, a graded approach is only used in determining the scope and level of detail of the safety assessment carried out in a particular state for any particular facility or activity. In this document, the graded approach is applied for many applications of ETRR-2 research reactor, such as; the QA level determination during ETRR-2 fabrication and construction stages, and the frequency of Inspection, periodic testing and maintenance. Each application is graded quantitatively, which is considered to be a new trend. Grading was applied based on many factors such as; safety, reliability, design state, complexity, experience, availability, and economic factors. A certain amount of points were assigned to each factor. A formula then be applied to obtain the total amount of points. This total rating may correspond to a general system or to its components. Key Words: Graded approach, ETRR-2, Maintenance, Quality assurance

Speaker: Dr Yasser ELLETHY (ELLETHY)

Paper Graded_approach__paper__.pdf

Slides 27_GRADED_APPROACH_paper.ppt

11:50 Safety Reviews of Research Reactors in Germany – Graded Approach for the periodic safety review according to § 19a of the Atomic Energy Act

In Germany, a total of 46 research reactors were built and operated. In the meanwhile, most of them are in decommissioning or have already been dismantled completely. At present, only 7 research reactors are still being in operation. They include: • 2 large pool reactors: FRM II in Garching near Munich with a thermal output of 20 megawatts (MW) and BER II in Berlin with a thermal output of 10 MW. • 1 TRIGA Mark II reactor in Mainz with a thermal output of 100 kilowatts (kW). • 4 smaller training reactors, the so-called zero-power reactors: AKR-2 in Dresden with a thermal output of 2 watts and 3 SURs (Siemens training reactors) in Stuttgart, Furtwangen and Ulm with a thermal output of 100 milliwatts (mW) each. These research reactors differ from each other not only with respect to their design, thermal power and radioactive inventory but also with respect to the nuclear fuel used, the mode of operation as well as to the site (e.g. central location in a city or in a suburb). Consequently their risk potential is also very different. To assure the safety of German nuclear installations, a comprehensive national legislative and regulatory framework has been established. The most important document is the Atomic Energy Act [1] together with its associated ordinances. They constitute the legal basis and are directly binding to all kind of nuclear installations in a common approach, including research reactors. Within this national legal framework, the licencee has the prime responsibility for the safe operation of its own nuclear facility. To ensure safety over the entire lifetime of the facility, the operator is obliged among the other to perform particular safety reviews: • within the licensing procedure according to the paragraph 7 of the Atomic Energy Act, i.e. in case of construction, operation, essential modifications of the installation or its operation as well as for decommissioning of the facility • safety upgrades, which are carried out within the continuous regulatory supervision pursuant to paragraph 19 of the Atomic Energy Act. Moreover, according to the paragraph 19a of the Atomic Energy Act, which was introduced in its 12th amendment in 2010: “(3) Anyone who operates any nuclear installation […] shall perform a verification and evaluation of the nuclear safety of the respective installation every ten years and shall improve nuclear safety of the installation continuously. The results of the verification and evaluation shall be submitted to the supervisory authority. (4) The evaluation according to para. (1) or (3) shall encompass the verification that measures are taken to prevent accidents and to attenuate the effects of accidents including the verification of the physical barriers as well as of the administrative preventions of the licencee which would have to fail before life, health and material assets are damaged by the effect of ionising radiation. The competent supervisory authority can issue orders concerning the extent of the verification and evaluation by the licencee.“ This means, that the regular safety reviews similar to the Periodic Safety Reviews for nuclear power plants become mandatory for all kind of nuclear facilities, which includes also research reactors. The focus of the paper presented here is the evaluation of the risk potential of individual research reactor facilities and the development of an appropriate graded approach to perform the safety reviews according to the paragraph 19a of the Atomic Energy Act. [1] Act on the Peaceful Utilisation of Atomic Energy and the Protection against its Hazards (Atomic Energy Act) of 23 December 1959, as amended and promulgated on 15 July 1985, last amendment by the Act on 28 August 2013, BfS Safety Codes and Guides, Translations Edition 08/13

Speaker: Dr Katarzyna Niedzwiedz (Federal Office for Radiation Protection)

Paper IAEA-CN231-64_Paper_K.Niedzwiedz.pdf

12:10 Leveraging Safety Programs to Improve and Support Security Programs

There has been a long history of considering Safety, Security, and Safeguards (3S) as three functions that need to be integrated along with operations into an effective physical protection system. Here, we look instead at how safety programs can be extended to directly benefit security. Our focus will be on nuclear facilities but similar ideas could be used to support security programs at other types of high-consequence facilities. There are several inherent advantages of safety programs over security programs, such as: - Safety, as a field, is very mature and has very detailed processes; - Safety processes and techniques can be shared nationally and internationally without security concerns; - Nuclear facilities, in particular, tend to have fairly extensive safety programs. By having a mature and robust safety program, there are also intangible benefits to security performance relative to organizational trust, identification of suspect behaviors and/or threats, and mature organizational communication for appropriate response to security related events. It is our contention that it would be better to build on a site’s nuclear safety program to leverage these strengths over into the security area than to build similar capabilities from scratch within security organizations. We believe that the strengths of the safety program can be carried over to security in a relative straightforward and cost-effective manner. The resulting capabilities could then be demonstrated very quickly to the international community at Sandia National Laboratories’ Integrated Training Facility. This paper reflects discussions at Sandia involving reactor safety, operations, and engineering organizations as well as security experts. Note that Sandia operations and engineering staff have experience with both commercial nuclear power plants and research reactors and have existing work with the International Atomic Energy Agency (IAEA). The paper discusses several potential areas where safety programs and capabilities could benefit security: -Security Operations: to improve processes for developing and managing security and contingency plans, to include configuration management, for example, and to achieve better processes; - Quality Management programs: to include identifying measurable performance goals, determine performance requirements and to apply rigor to performance testing in such areas as test plan development, analysis, metrics, and reporting; - Security culture: to exploit the significant similarities between safety and security culture, for example ensuring that management takes a proactive role in assuring nuclear safety and security throughout an integrated organization by effective leadership, displaying accountability, proper resource management, and appropriate risk identification and mitigation. - Exercises where operations groups at nuclear facilities play a role during a security response: to illustrate the role that these groups can play during a security response. Reactor operations and emergency response groups could take part as role players as operators of simulated, notional nuclear facilities as part of simulated security exercises in order to demonstrate how safety, security, and emergency response organizations can work together effectively (and to reflect this knowledge in site plans). In particular there are two general safety- related programs (all intimately tied with security culture) that could be extended to benefit security: - Management systems:1 “The term management system reflects and includes the initial concept of ‘quality control’ (controlling the quality of products) and its evolution through quality assurance (the system to ensure the quality of products) and ‘quality management’ (the system to manage quality). The management system is a set of interrelated or interacting elements that establishes policies and objectives and which enables those objectives to be achieved in a safe, efficient and effective manner. “ - Formality of operations:2 This covers policies and practices to be followed by employees to ensure safety in the workplace, especially in hazardous industries. Formality of operations includes conduct of operations (including following well-defined procedures), configuration management, maintenance and surveillance, and training and qualification. Note, also, that the management system should support the enhancement and improvement of safety culture3. Other relevant IAEA documents also exist4. These IAEA documents can be found at: http://www.iaea.org/NuclearPower/ManagementSystems/ Finally, the paper discusses how Sandia’s Integrated Training Facility could be used to demonstrate how this safety/security integration could be accomplished.

Speakers: Ms Janice Leach (Sandia National Laboratories), Mr Mark Snell (SNL)

Paper 154_Leveraging_Safety_Programs_to_Improve_and_Support_Security_Programs_Revised_Format.pdf

Slides 154_Leveraging_Safety_Programs.pptx

12:30 → 14:00 Summary, Recommendations and Closing Remarks

Convener: Mr Pablo ADELFANG

15:00 → 17:00 Technical Tour: Atominstitut