Powered by Indico
v3.2.9
Since 18 of December 2019 conferences.iaea.org uses Nucleus credentials. Visit our help pages for information on how to Register and Sign-in using Nucleus.
The safe, secure, reliable and economic management of spent fuel arising from nuclear power reactors is key for the sustainable utilization of nuclear energy and covers many technological aspects related to the storage, transportation, recycling and disposal of the spent fuel and the high level waste (HLW) generated from spent fuel reprocessing.
The sustainability of nuclear energy involves the preservation of natural resources and the minimisation of generated wastes. Currently, in some countries, the remaining uranium and plutonium are industrially recovered from spent fuel and recycled as mixed oxide (MOX) fuel in thermal reactors, saving natural uranium resources and generating vitrified HLW and irradiated spent MOX fuel. Future advanced fuel cycles based on the implementation of Gen-IV reactors will allow nuclear energy to be almost independent of uranium natural resources in addition to dramatically reducing the generated wastes in terms of heat loading, radiotoxicity and proliferation risks.
The last IAEA International Conference on Management of Spent Fuel from Nuclear Power Reactors, held in June 2015, highlighted that there is little integration in the fuel cycle in terms of analysing how decisions made in one part of the fuel cycle may affect another part. Introducing efficiencies into the individual steps in isolation can create additional challenges in subsequent steps. A worst-case scenario would be that a decision taken today forecloses a transition to another step tomorrow. In most cases a technical solution can be found, but this is likely to come at a price (cost, resource utilization, etc.). Therefore, one of the main challenges is to maintain enough flexibility to accommodate the range of potential future options for the management of spent fuel as well as to define and address the relevant issues in storage and transportation, given the current uncertainties regarding the storage duration, the availability of future technologies and future financial, regulatory and political conditions.
In this context, the IAEA is organizing the International Conference on the Management of Spent Fuel from Power Reactors in Vienna from 24 to 28 June 2019, with the theme “Learning from the Past, Enabling the Future”.
The scope of the conference covers the management of spent fuel from nuclear power reactors from the past, present and future nuclear energy systems, and how it can be affected by the decisions taken in the rest of the nuclear fuel cycle.
The management of the spent fuel in the EU is addressed in alignment with Council Directive 2011/70/Euratom, which aims at the safe and responsible management of radioactive waste and spent fuel in order to avoid imposing undue burdens to the future generations; at ensuring the highest levels of safety; and at ensuring transparency and the involvement of the public in the decision-making process. Twenty one EU Member States manage about 59 000 tHM of spent fuel generated in past and current nuclear power generation and nuclear research activities. Each year, about 3 200 tHM of additional spent fuel are generated. Some Member States reprocess spent fuel and some others have decided to keep this option open. The majority of the EU Member States have opted for direct disposal of their spent fuel. Right now the EU does not have in its territory any facility for the disposal of spent fuel, high level and long-lived radioactive waste. Finland, Sweden and France expect to start the operation of their deep geological disposal facilities within the next two decades, while the rest of the MS with nuclear programmes have planned operating disposal facilities in the time interval 2040-2130, with a peak in the decade of 2060-2070. Long-term or extended interim storage is thus instrumental in the national strategies for the management of spent fuel prior to reprocessing or disposal. The Euratom Research and Training Programme contributes, within its portfolio of activities, to the safe management of spent fuel and radioactive waste. This is done through indirect research and innovation activities to which the European Union provides financial support and which are undertaken by EU Member States research entities, and through direct research and innovation activities undertaken by the Commission through its Joint Research Centre (the ‘JRC’: the European Commission's science and knowledge service). This paper provides an overview of the JRC areas of research relevant for safety of spent fuel (and high level radioactive waste), which cover all stages of spent fuel management since it is removed from the reactor: cooling in the spent fuel pool; handling, transport, storage (with particular emphasis on long-term storage); retrieval, handling and transportation after storage; disposal in a deep geological formation, and long term safety aspects thereafter. The paper highlights the main achievements, and the main challenges, stressing the relevance of the experimental work carried out on "real" spent fuel in JRC's research infrastructure, which include hot cells and other shielded facilities that are relatively rare or even unique.
Kenya is considering the development of a nuclear power programme for electricity generation in order to address the increasing national energy needs. The programme is currently in the second phase of implementation in accordance with the International Atomic Energy Agency (IAEA) milestone approach. One of the major infrastructure issues being developed is radioactive waste management. Currently, the bulk of radioactive waste generated in Kenya is disused sealed and unsealed radioactive sources. Small volumes of additional waste are also generated through the use of radioactive sources in medicine, industry, and research. However, with the planned introduction of nuclear power in Kenya, the radioactive waste from nuclear power generation will be orders of magnitude higher (in quantity, toxicity and half-life) than the waste generated from medical, research, and industrial practices. A national policy and strategy for managing radioactive waste and spent fuel is important as it sets out the nationally agreed position and/or plan for managing spent fuel and radioactive waste. It is also a visible evidence of the concern and intent of the government and the relevant national organizations to ensure that radioactive waste and spent fuel are properly taken care of. The absence of explicit policies and strategies can result in lack of transparency and accountability. There is, therefore, a need to develop a robust policy and strategy for the safe, secure and safeguard-able management of radioactive waste and spent fuel. The proposed national policy and strategy framework for Kenya takes into account a variety of technical and managerial elements/aspects that will ensure the radioactive waste, including spent fuel, from the nuclear power plants, is managed sustainably in the long-term so as to avoid imposing an undue burden on future generations.
Indonesia has been acquainted with nuclear energy utilization since the late 1950s, when President Soekarno, intended to acquire positive benefit from nuclear energy. After that, Institute of Atomic Energy was established in 1959, which was renamed into Nuclear Energy Agency (BATAN). The first research reactor (TRIGA 2000 Reactor) was built in Bandung in 1961 and was operated in 1965. During President Soeharto regime, two other research reactors have been built, a 100 kilowatt Kartini Reactor in Yogyakarta (operated in 1979, and a 30 Megawatt Multipurpose Reactor, RSG Siwabessy in Serpong (operated in 1987). Until now, these three research reactors have been operating safely. The type of fuel of those reactor is Uranium Oxide for Kartini Reactor and TRIGA 2000 and Uranium Silicide for RSG Reactor. The operation from these three reactors produces approximately more than hundreds spent fuel or irradiated fuel. Based on the Government Regulation No. 61 Year 2013 on Radioactive Waste Management (GR 61/2013), the licensee should manage those spent fuel in the aspect of safety, security and safeguards. This Government Regulation is an implementation of some clauses in the Act No. 10 Year 1997 on Nuclear Energy. In this Act the spent fuel is classified as high level of radioactive waste management.Considering the commitment to this international convention and its consequences, Indonesia has developed infrastructures by strengthening legislation and regulation. In the near future, Indonesia is planning to build Nuclear Power Plants that will be owned and operated by private companies. Therefore, the Nuclear Energy Regulatory Agency (BAPETEN), as the nuclear regulator in Indonesia, must prepare to ensure this policy can also be implemented. This paper will examine the further challenges that will be faced by Indonesia, especially from BAPETEN’s perspective, in order to ensure the spent fuel management policy can be implemented properly.
The paper has the purpose to introduce technical, informative and economic aspects about the managing of spent nuclear fuel in Brazil. The capacity of spent nuclear fuel inside the spent nuclear fuel pools from Unit 1 and Unit 2 of Angra nuclear power plant is almost reached. Also, the content in it will show the reader the importance of managing spent fuel in order to extend the lifetime operation of nuclear power plants, since the context of the country energy matrix until the technical solution found. The arguments presented are based on technical publications from different institutions. It aims to present the reader the strategy founded by the Brazilian nuclear power plant operator in order to meet the requirements to renewal its operational license addressed by the regulatory body.
Strategic analysis of the principal spent nuclear fuel management approaches is reported as they are seen for Belarus. Particular features of open and closed fuel cycles are considered and compared. The deferred decision perspectives for the nuclear fuel cycle back-end are discussed. Few available and arising options for the back-end are analyzed taking into account modern trends and technology developments in Russian Federation as the principal supplier of nuclear technologies and in other countries. Some results of feasibility evaluations of long term spent nuclear fuel storage are presented. General requirements for the storage system are formulated. Problems and perspectives concerning the reprocessing of high burnup fuel are presented in the context related to Belarussian NPP. The topic of high level waste management arising after nuclear fuel reprocessing as well as the perspectives of direct spent fuel disposal in Belarus are briefly discussed. Possible intermediate level radioactive waste management strategies including its disposal are considered in some details. Available strategic approaches for the spent fuel management in Belarus are outlined and compared. Recommendations for the national strategy and the short-term national action plan are given.
TThe Interim Spent Fuel Storage Facility (ISFSF) Jaslovské Bohunice purpose was to cool down the spent fuel under water before its transport to the Soviet Union. After the political change this transport became impossible. Therefore, the capacity of ISFSF had to be increased. However, the storage capacity will be fully exhausted in 2022. As there is not any other option yet, the solution is to expand the current storage capacity.
The operator of ISFSF, the Nuclear and Decommissioning Company (JAVYS), decided to build a new dry storage. In 2016 the EIA process was successfully completed and after selection of contractor in 2018 the preparatory work started.
An important role in this process has been performed by the Nuclear Regulatory Authority of the Slovak Republic (UJD). After extensive discussions the basic requirements have been set up. New storage capacity will be dry, vault type storage with passive residual heat removal.
At present the basic documents are completed. The Technical Report "Assignment to Safety Analyzes" was approved, and the technical report will, in addition to the assessment of normal operation and its impact on the radiation protection of the population, also assess the radiation risk to the population in the following postulated internal initiation events: loss of power supply, transport and handling accidents, accidents associated with loss of heat removal, criticality accidents, loss of the monitoring functions, the leakage of primary and of the secondary lid of the canister. External initiation events: explosion, fire, flooding, earthquake, extremely hot and cold and plane crash.
The UJD will pay particular attention to requirements for seismic resistance and protection against accidental fall, or deliberate attack with airliner. UJD will also demand all spent fuel suitable for dry storage to be transported form wet to dry store. This will increase nuclear safety and nuclear security as well.
The paper covers the Spanish national policy and strategy for spent nuclear fuel, high level and special radioactive waste management. The existing legal framework in Spain establishes the need to keep a General Radioactive Waste Plan (GRWP) up to date. The basic strategy for the management of the SF, HLW and SW aims for their future disposal in a deep geological repository (DGR). Such stage will be preceded by a temporary storage in a centralized facility (CSF). As this is not yet available, some actions have been performed in the NPPs to avoid the saturation of spent fuel storage pools and to allow, in this way, that they could either continue to operate or to be dismantled.
The paper also describes the main processes of the CSF, its design criteria, safety case, and site selection procedure.
As part of its efforts to help resolve the major climate and energy issues facing future generations over the next decades, France is committed to a global energy transition materialised through the Act of 17 August 2015 on the energy transition for green growth (LTECV). This act defines the main objectives for the medium and long term. Among these objectives, it is worth highlighting:
- Reduction in greenhouse gas emissions by 40% between 1990 and 2030, and a 4-fold reduction in greenhouse gas emissions between 1990 and 2050
- Development of renewable energy sources to reach 23% of the gross final energy consumption in 2020 and then 32% in 2030.
- Reduction in nuclear energy’s contribution to electricity generation to reach 50% by around 2035.
To achieve these objectives, the LTECV Act specifies the definition of a French national strategy to lower carbon emissions (SNBC) and a multi-year energy programme (PPE). The first version of this programme covers the periods 2016 to 2018 and 2018 to 2023. It must be reviewed every 5 years over a 10-year period. The main orientations of this PPE programme for the 2019-2028 period were published by the French government within the scope of a project announced in January 2019; they will be open to public consultation before their adoption scheduled for the end of summer.
For nuclear power to be a sustainable option for Sudan, depleted nuclear fuel management techniques must be implemented that meet strict safety and environmental protection standards, so one solution to these problems may be to use transmutation to convert the nuclides in spent nuclear fuel to ones with shorter half-lives. Both reactor and accelerator-based systems have been examined in the past for transmutation;and development of fast reactors that can also burn the minor actinides recycled from spent fuel through applied research with neutron of the National Center for Nuclear and Radiological Control. This study examines technical issues, institutional factors and strategic options for managing spent nuclear fuel, and draws on policy implications and those associated with different social priorities and values.
The management of spent fuel from nuclear power reactors still pose a great challenge with varying competing management and disposal strategies. It is indeed a challenge to talk about the management of spent fuel without considering the main issues of its association with nuclear weapons, and the bitter legacy left by the major nuclear accidents at Chernobyl and Fukushima. However, it is encouraging to know that the lessons learned from the past have motivated scientists and engineers to design modern nuclear reactors with systems to maintain cooling in case the primary coolant is lost. This extended abstract presents the management of spent fuel and the role of the National Radioactive Waste Disposal Institute (NRWDI) in South Africa.
The one third of power generation in Armenia is from the Armenian Nuclear Power Plant (ANPP). Currently is under operation only second unit of ANPP. The safety of the ANPP is a top priority for the Government of Armenia. Strategy on the Safe Management of Radioactive Waste and Spent Nuclear Fuel and the Action Plan on its implementation was approved by Government Decrees in 2017 and 2019. The source of spent fuel is the ANPP. When its service life is expired, during refueling of the reactor core, fuel assembly is discharged from the reactor core and placed in cell of the Unit № 2 storage pool. Refueling is performed once in a year, when the reactor is shut down, depressurized and cooled down. Following discharge of spent fuel assembly from the reactor core it is tested for leak tightness. In case it is tight, a spent fuel assembly is placed in a cell of the storage pool. In case failed assemblies are detected they are placed in tight casings and stored in them. Following the required storage time (3-5 years), spent fuel assemblies are relocated into the spent fuel storage pool of № 1, where they are stored until reaching the parameters required for their 2 transferring to Dry Spent Nuclear Fuel Storage facility (DSNFSF) of NUHOMS-56 type located on Armenian NPP site. Spent fuel assemblies are placed in Dry Shielding Canister (DSC), which is filled with nuclear purity helium of retained overpressure. The DSC shielding and insulation are provided by massive reinforced concrete Horizontal Storage Module (HSM). Radioactive decay heat caused by DSC and HSM is removed using draught ventilation system, which operation is based on a passive natural convection. There are two buildings of the HSM built on the Armenian NPP site
The poster report presents the structure of the Strategy of spent fuel management to NPP in the Republic of Belarus. A brief overview of the documents required for the development of the Strategy is presented, as well as relevant intergovernmental agreements.
The presentation provides the different approaches are being considered for spent fuel management, including longtime storage of spent fuel in the Republic of Belarus (the construction of a "dry" store) and sending of the spent fuel to the Russian Federation for temporary technological storage with further reprocessing and the final disposal of radioactive waste in the Republic of Belarus.
The creation of necessary infrastructure of spent fuel management, the design and construction of a temporary spent nuclear fuel storage and final disposal of radioactive waste, define supply chain for the export of the spent fuel and return of conditioned radioactive waste for final disposal in Belarus is presented.
The presentation also provides the information on the cooperation with the IAEA. The progress of the construction of the first NPP in the Republic of Belarus is presented in the report.
India follows closed fuel cycle option for spent fuel management. Wet storage of spent fuel is the predominant mode of storage therefore the discharged fuel from the reactors is stored at the reactor pools which have capacity for ~10 reactor-years of operation. After appropriate cooling, the spent fuel is moved to the storage locations either on or off reactor site depending on the spent fuel management strategy. Transport of the spent fuel is carried out adhering to national and international safety guide lines in "type B" packages. Lower capacity fuel ponds are provided for interim storage of spent fuels at recycling facilities. PUREX process using TBP is employed for reprocessing spent fuel from PHWRs. Spent fuel reprocessing from FBRs and futuristic reactors is demonstrated using TBP based solvent extraction processes. The safe management of radioactive wastes envisages two distinct modes of final disposal in respect of radioactive wastes viz. near-surface engineered, extended storage for low and intermediate level radioactive wastes and deep geological disposal for high-level and alpha bearing wastes. HLLW treatment is carried out in waste immobilization plants and interim storage of vitrified HLLW is carried out in solid storage and surveillance facilities. Extensive R&D in partitioning of long-lived actinides and fission products has lead to the development of solvent extraction based process flow-sheets using indigenously synthesized solvents which are deployed at engineering scale. This has resulted in the reduction of waste volume generation and extended time of repository requirements. This has also resulted in the recovery of several useful radionuclides such as 137Cs, 90Sr, 106Ru etc. which are used for societal benefits.
To develop nuclear energy is inevitable choice for China to meet the requirement of decreasing greenhouse gas emission，at the same time of economic and society development. To ensure sustainable development of nuclear energy, closed nuclear fuel cycle strategy based on fast reactor has to be adopted. Both of recent and next R&D activities of nuclear fuel cycle back-end were introduced in the paper, such as:
- Nuclear energy development and spent fuel accumulation, including fast reactor and ADS development aiming at transmutation long-lived nuclides.
- Commissioning of Reprocessing Pilot Plant for PWR spent fuel, development of advanced PUREX process and hot test of separation both U and Pu in CRARL (China reprocessing and radiochemistry laboratory).
- Minor Actinides separation on laboratory scale.
- Investigation on vitrification of high-level liquid waste, high level waste disposal and its programme.
Thirty years of watching attempts at implementation of a U.S. national strategy for high level waste management embodied in the Nuclear Waste Policy Act and its Amendments (of 1982 and 1987) from many vantage points have led to strong personal views on what has gone wrong with U.S. strategies. Instead of a repository open in 1998, the U.S. is still probably at least two decades away from opening a repository. My vantage points include management of the Los Alamos National Laboratory research programs for Yucca Mountain, years on the staff of the U.S. Senate, Commissioner of the Nuclear Regulatory Commission, and Assistant Secretary responsible for implementation of these strategies. In the talk, the stark differences between the path followed so far by the U.S. and the path recommended by the U.S. President's Blue Ribbon Commission on America's Nuclear Future will be discussed.
The basic policy of Japan is to promote a nuclear fuel cycle that reprocesses spent fuels and effectively utilizes the plutonium etc. retrieved, from the viewpoint of effective utilization of resources and reduction of the volume and harmfulness of high-level radioactive waste. This paper explains our policy and efforts regarding the nuclear fuel cycle.
45 years since it was first conceived and after reprocessing over 9300 tonnes of fuel, THORP sheared its last fuel assembly on 9th November 2018. Providing a vital service to UK, European and international reactor operations, the facility will continue to store fuel for at least the next 50 years. This presentation will look back at some of the history, the economics and lessons learnt more than 24 years of successful operation and forward to a new future for the staff and facility.
At present, Russia's nuclear power industry continues its development and increases its contribution to the overall energy mix, which reached 18.9% in 2017. The basis of nuclear power generation is formed by LWR, in the same time Russia operates two industrial-size fast reactors – BN-600 and BN-800. It is expected that from the year 2030 there will be the large-scale implementation of fast neutron power reactors and the transition to a two-component nuclear system with unified fuel cycle, linking the needs of both existing thermal reactors and fast neutron reactors. Solving the problems associated with the accumulation of SNF and radioactive waste in this regard is becoming a priority.
As a basic approach to SNF management in Russia, the concept of its reprocessing with the nuclear materials recycling in a two-component nuclear power energy system (with thermal and fast neutron reactors) has been adopted. The main purposes are an efficient use of natural uranium resources, SNF non - accumulation, recycling nuclear materials, and reducing the radiotoxicity and volume of the generated radioactive waste.
Russia has many years of experience in safe management of spent nuclear fuel from power reactors including storage, reprocessing and recycling . The reprocessing plant RT-1 has been operating since 1977. To date, over 6,000 tons of SNF have been reprocessed. At the same time, a integrated complex for SNF management is being created at the site of the Mining Chemical Combine, which includes: centralized water cooled (“wet”) SNF storage; centralized air-cooled (“dry”) SNF storage; a pilot-demonstration centre for the reprocessing of SNF based on innovative technologies; MOX fuel fabrication for fast neutron reactors (BN-800 type). An underground research laboratory will be set up here to develop the technologies for the HLW final isolation.
The recycling of repU is currently being fully implemented during the fabrication of fuel for thermal reactors. Separated plutonium from LWR SNF starts involving in NFC as a component of MOX fuel for FR (for starting loading and feeding during the first 10 years of operation of fast reactors). At the same time, the technology of multi-recycling in thermal reactors of plutonium and repU from LWR SNF is being developed (REMIX-concept).
To reduce radiotoxicity and the volume of ultimate wastes to be disposed of, HLW partitioning technologies are being developed with MA and heat-generating fission products recovering. Russia already has industrial experience in HLW partitioning.
The technology of MA transmutation is planned for studying using both solid-fuel fast reactors (like BN-800 type) and MSR.
Canadian Nuclear Laboratories (CNL) in Chalk River, Canada has two experimental long term storage programs for spent nuclear fuel: wet storage and dry storage. The objective of both programs is to determine the length of time spent nuclear fuel can be safely stored in specific environmental conditions and to characterize the condition of the fuel as a function of time via periodic examinations. Both programs were initiated in the 1970’s and have stored fuel in wet and dry conditions in excess of 50 years and 40 years, respectively. This paper discusses the history of CNL’s long term storage programs, a summary of previous results, and preliminary plans for continued experimentation commencing in 2019.
In order to alleviate the pressure brought by the rapid growth of spent fuel in NPPs, insufficient reprocessing capacity and high construction cost, China has carried out the strategy of technology import and equipment localization to construct the PWR spent fuel dry storage project for the first time. The paper firstly introduces the project implementation strategy, product selection and bidding, engineering construction. Then it presents the main experience feedback which includes: 1) Taking full advantages of Architecture Engineering model widely used and verified in NPPs construction in China, promoting the coordinated progress of spent fuel dry storage project design, construction and domestic dry storage components manufacturing industry chain; 2) Based on its own needs, carrying out comprehensive investigations and studies before importing the technology, fully identifying the import risks and formulating plans in time; 3) Clarifying the responsibilities of all parties in the contract, and accurately defining the boundary of intellectual property rights; 4) Actively cooperating with the regulatory authorities to fill the gaps in domestic regulations and standards. In addition, according to the reprocessing strategy and the demands of NPPs customers, China has independently researched and developed spent fuel dry storage components, which has formed independent intellectual property rights. And the total price of spent fuel canister is reduced by about 53% compared with imported equipment, which has significant economic benefits. Finally, in view of the shortcomings of the existing dry storage technology, the paper also prospects the future technological development, and proposes a variety of new product design schemes for the first time, which aims at jointly promoting the technological progress of the spent fuel dry storage and sustainable development of nuclear energy.
Clab is a facility, own by Swedish Nuclear Fuel and Waste Management Co (SKB), for wet intermediate storage of spent nuclear fuel pending deposit in final repository. At Clab 6700 tonnes of nuclear fuel are currently stored with a residual power of about 8.3 MW. Requirements regarding ageing management programs at nuclear facilities were introduced in 2006. Two attempts to introduce ageing management programs were made between 2006 and 2013, but failed. In 2013, SKB received an injunction to implement an ageing management program for Clab.
A project group was appointed to produce an appropriate ageing management program for the facility. The program was developed with guidance from the IAEA safety guide No NS-G-2.12.
The strategy was to involve the line organization early in the project so that the people who would manage the program were involved in the development of management systems, analyses and proposals for measures. This made the handover of the project to the line organization quite simple and straight forward.
The facility consists of approximately 160 systems, of which about 96 are included in the ageing program. Only systems that are important for radiation safety are currently included in the program.
After all systems were analysed from an ageing perspective, 546 new measures, were identified that needed to be implemented to have control over the facility’s ageing. During the execution of the measures, several unexpected discoveries have been made.
The result of the work in the program has shown that the plant's status with regard to physical ageing is good. Technological ageing (obsolescence) is a bigger challenge.
Several lessons were learned in the development of the ageing program, for example the importance of good communication with the supervisory authority. Another lesson is the importance to set the right level of analysis that otherwise risks becoming ineffective.
The paper describes the strategy adopted by Sellafield Ltd for management of the remaining lifetime arisings of AGR fuel from EDFE reactors. AGR reprocessing operations have completed at Sellafield but fuel will continue to be received, dismantled and consolidated in line with current practice. Spent fuel will be wet stored in existing facilities for an interim period until a disposal facility becomes available, extending fuel storage time from the current 5 years (for buffer storage pending reprocessing) to 80 years. The main safety issues associated with this interim storage strategy are ensuring the long term integrity of the fuel to be stored and the structure of the storage facility. The paper summarises research on fuel corrosion resistance and the pond structure inspection reports. Storage for the interim period requires changes to the operations of facilities and examples of these changes are given.
Neutron absorber materials (NAMs) are used in spent fuel pools (SFPs) to maintain criticality safety margins while increasing fuel storage space. SFP lifetimes are increasing, and operating experience has documented that there are a number of pools without a coupon monitoring program or with a limited number of coupon samples remaining. To address the long-term monitoring needs globally across the industry, EPRI has initiated the development of the industrywide learning aging management program (i-LAMP) for NAM monitoring in SFPs. This program is initially focused on BORAL®, the most widely used material in SFPs—especially in the United States—and will later be extended to other metallic neutron absorber materials. In addition to participation by all U.S. utilities, several other countries (for example, Mexico, South Korea, and Taiwan) are participating in the program; the aim is to increase global participation to achieve a globally-applicable program. As part of the program, SFP water chemistries and coupon analysis results to date are being collected. From these data, analysis is performed to determine additional data needs as well as analysis for the development of the sister pool criteria and learning aging management program. The program will also allow trending, timely identification of outliers and any potential concerns, and development of an improved technical basis for guidelines and future monitoring. The paper presents the proposed i-LAMP, the components of the i-LAMP, data collected to date, and a roadmap for the development and implementation of the i-LAMP.
The Federal Office for the Safety of Nuclear Waste Management (BfE) is the competent licensing authority for interim storage of spent nuclear fuel (SNF) and high-level radioactive waste in Germany. The concept of dry interim storage, comprises dual purpose casks equipped with a double barrier lid system with permanent monitoring of its leak-tightness.
Existing storage licences in Germany are limited to 40 years. Due to the time needed for site selection, construction and commissioning of a deep geological repository, a prolongation of the interim storage period will be necessary to bridge the gap until final disposal. To demonstrate if safety requirements could be fulfilled by the transport and storage cask beyond the initially licensed 40 years additional research is required.
Research towards material degradation e.g ageing of cask materials or internals, fuel assembly behaviour and behaviour of storage facility buildings and operational equipment is the basis for the safety of SNF storage. To identify potential fields of further interest it is also necessary to acquire additional data for the above mentioned research.
As interim storage facilities are a key step towards the final disposal, it is also necessary to conduct research towards the impact of prolonged interim storage on the final disposal, especially due to the increasing relevance of ageing effects like material degradation. This includes foremost data acquisition and storage to enable a safety based choice of actions.
Furthermore, as Germany is phasing out of nuclear energy the knowledge management in the nuclear field gains enormous importance especially in regard of human resources.
The BfE as licensing authority has initiated several research projects to cover the foresaid topics to be presented in this article.
Aging management is applied to dry storage systems used to store used nuclear fuel to ensure that material degradation does not affect the function and safety of these systems as they remain in service beyond the initial licensing period. Chloride-Induced Stress Corrosion Cracking (CISCC) is a potential degradation mechanism for welded stainless steel canisters that serve structural and confinement functions in some dry storage system designs. EPRI has developed aging management guidelines to address the potential for CISCC in these canisters. The guidelines include recommendations for screening and inspection methods and frequency with a technical basis built on literature survey results, qualitative failure modes and effects analyses, deterministic flaw growth and tolerance calculations, susceptibility assessments, and probabilistic canister confinement integrity assessments. EPRI’s work is being referenced by the American Society of Mechanical Engineers (ASME) Boiler Pressure Vessel (BPV) Section XI Task Group on In-service Inspection of Spent Fuel Storage and Transportation Containments which was formed in April of 2015.
At present, Ukraine has two spent fuel storage facilities under operation: wet spent fuel storage at Chernobyl NPP for RBMK-1000 fuel and dry spent fuel storage at Zaporizhzhya NPP for VVER-1000 fuel. Two more dry spent fuel storage facilities are under construction. The design of the ZNPP dry spent fuel storage facility was based on the VSC-24 interim spent fuel storage system used in the United States of America. Due to differences in the US and Ukrainian nuclear fuel, the loading of the ZNPP dry spent fuel storage facility could not be justified without additional measures. Some features of the wet spent fuel storage facility needed justification as well. This report presents the main approaches that were used in justification of the wet and dry spent fuel storages in Ukraine, as well as their state technical reviews.
The evaluation of cladding integrity is a major issue to be demonstrated in Germany for extended interim storage periods up to 100 years and subsequent transportation considering operational and accidental conditions with respect to reactor operation, cask drying and dry interim storage. The chemical reaction between the zirconium fuel cladding and the cooling water in water-cooled reactors produces hydrogen and zirconium oxide. Hydrogen diffuses into the cladding and precipitates as zirconium hydrides when the solubility limit is reached, preferably oriented in hoop direction. At high temperatures during vacuum drying procedures, the hydrides can dissolve. Over a succeeding period of slow cooling with existing hoop stress the hydrides precipitate again, but partly reoriented along the radial direction of the cladding. This change of microstructure in combination with a decreasing temperature (0.5...2 K/year) during (extended) interim storage and additional mechanical load by handling procedures or under accident conditions could lead to a potential cladding embrittlement and consequently increased failure probability. The current research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) has been launched by BAM to investigate potential spontaneous brittle failure of spent fuel claddings at small deformation under long-term dry interim storage conditions. Based on the key thesis that radial hydrides may be considered as sharp cracks, BAM plans Ring Compression Tests (RCT) with unirradiated cladding samples with representative hydride distribution (including hydride reorientation), numerical simulation of the RCT, calculation of fracture toughness, and identification of failure criteria. Without hydride reorientation, samples under RCT conditions show large plastic deformation with gradually decreasing force at the end of the test indicating ductile failure. Contrary, with hydride reorientation, spontaneous failure with abruptly decreasing force at very small deformation and low temperature is possible dependent on hydrogen content and mechanical load during hydride reorientation.
In order to use the concrete cask system for the storage of spent fuels in Japan, it’s necessary to establish method to manage aging effect, especially SCC (Stress Corrosion Cracking). Therefore, the methods are under development to mitigate SCC, to inspect crack, and to monitor confinement of a canister. For the mitigation of SCC, one of following factors should be kept in the certain conditions; residual stress, environment and material. The method has been chosen to make surface residual stress compressive by ZSP (Zirconium Shot Peening) and WJP (Water Jet Peening). At a fabrication factory, ZSP is used because of its economic advantage. At a nuclear power plant, WJP is utilized because its garbage, water is easy to treat at a nuclear power plant. The value of compressive residual stress and the depth of its layer induced by ZSP and WJP have been experimentally confirmed. Pitting corrosion can occur on a metal surface regardless of its conditions. Once pitting corrosion penetrates the compressive residual stress layer, SCC may occur at the tip of pitting corrosion and finally penetrates the canister shell. The pitting corrosion depth on SS Type 316L surfaces has also been measured for 10,000 hours under various conditions, and estimated its growth rate. The estimated depth of pitting corrosion during the storage period is shallower than that of the compressive residual stress layer induced by ZSP and WJP. Additionally, the development of Non-Destructive Testing (NDT) method to inspect cracks in case SCC penetrates the compressive residual stress layer has been succeed. As the canisters is loaded with spent fuels, the method should be available in radiation environment. Eddy Current Testing (ECT) is chosen from various NDT methods because it is easy to operate ECT automatically and remotely. The magnetized weldment of a canister made of SS Type 316L interferes with inspection of cracks with ECT. It has been succeed in improving ECT method which is able to remove the magnetic noises. This improved ECT method allows to detect cracks by a millimetre unit.
Relating to the method to monitor confinement of a canister, the principle of this method is that it’s only necessary to check the temperature difference between top and bottom of the canister in order to monitor the inner gas leakage. The principle has been confirmed experimentally and analytically. As mentioned above, the authors have established the methods to mitigate SCC, to inspect crack, and to monitor confinement of a canister.
The main function of the Interim Storage for Spent Nuclear Fuel (ISSF) in Serpong site is to store-underwater the Spent Nuclear Fuel (SNF) arising from the operation of the GA Siwabessy Multipurpose Reactor (GAS-MPR). Structure, system, and components (SSCs) of ISSF was built in 1997 and was designed by AEA Engineering, United Kingdom (UK - AEA). ISSF began to obtain operating licenses in 2008 which is valid for 10 years until 2018. In order to extend the operating licence for next decade, the operatory body have to ensure that the ISSF will be safely operated by determining the result of the facility performance and aging review. Review of the facility performance require some analysis of SSCs from nuclear installations to assess and concludes whether the SSCs would be working properly to support the function of ISSF or not. While the aging review should present determination of the latest performances and conditions of the SSCs, including evaluations of each age related to the failure or indication of significant material degradation and justification of performance, future aging process, and remaining operating life of components. The scope of this research includes review of facility performance and aging of the ISSF’s SSCs. The review was conducted major in SSCs which important to safety such as liner inspection by ultrasonic detector, sipping test of the SNF to detect leakage of the cladding, corrosion study of cladding and liner material, support system perfomance, etc. The results show that the thickness of the liner is still 3 mm, no leakage detected of the cladding which have tested, corrosion rate is 0.002 mpy which is predicted to hold for more than 40 years operation, the water parameters are well maintained, and all the support systems are working properly as designed. From the results of the facility performance and aging review, it can be concluded that ISSF still can be used to store the spent nuclear fuel safely.
Spent fuel information are essential to make a national policy for spent fuel management, to evaluate the safety of transportation, storage facility and disposal facility. For that reason, The AMORES program (Automatic Multi-batch ORIGEN Runner for Evaluation of Spent fuel) was developed and used to evaluate inventory, radioactivity, and thermal power of transport cask or storage cask. This code is very useful to evaluate the present and future spent fuel characteristic to provide fundamental data for informed decision-making at various stages of SNF management (storage, transportation, and disposal) by using the whole spent fuel data from 1978 to 2015 in Korea. The aim of the study is to expand the function of AMORES code for evaluating the safety of transport cask or storage cask. For this purpose, AMORES code can contain the cask specification and material information as a database in advance. This data can be modified as an input file of MCNP code or KENO VI for calculating radiation shielding and criticality automatically by AMORES code, respectively. Therefore, it can call the MCNP code or KENO VI, execute these code, and extract the results from output file form.
In the study, in order to AMORES code validation ,criticality evaluation of KORAD 21 dual purpose cask was performed using the cask specification and material information of this cask that was developed to KORAD (KOrea RADioactive waste agency).
According to the nuclear safety act of Korea, the effective neutron multiplication factor, keff including all biases and uncertainties at a 95 percent confidence level, should not exceed 0.95 under all credible normal, off-normal, and accident-level conditions. keff of KORAD 21 was evaluated as 0.3280 and 0.94132 under the normal condition and accident condition, respectively.
Indonesia has G.A. Siwabessy reactor (RSG-GAS) that is currently in operation. Indonesia also has Interim Storage for Spent Fuels (ISSF) that is store for Spent Fuels (SFs) of RSG-GAS 30 MWt. In the national policy on Radioactive Waste Management Strategy, SFs are categorized as High-Level Waste and are not recycled because Indonesia follows open nuclear fuel cycle. SF is a fuel element type for this Material Testing Reactor that has dimensions of 76.1 x 80.5 x 868 mm with contents 19.75 % of enriched U-235 and AlMg2 cladding 0.38 mm. After the use of the fuel element in the reactor, fission products such as noble gaseous: Kr-85m, Kr-87, Kr-88, Xe-135 and Xe-138 are generated and trapped in the matrix of the fuel element. The other fission products and Transuranic (TRU) elements may possibly leak to the environment if the isolation of the fuel cladding has a defect or crack. The fission products like Cs-137, I-133, and Sb-124 pass through the cladding by diffusion process. Currently, there are 245 SF elements remaining from the maximum storage capacity of 1448 SF elements in the ISSF. The sipping test is one of the appropriate methods to determine the integrity of SFs and to observe potential leaks of SFs cladding by detecting the existence of fission products. It has been identified that 20 SF elements releasing maximum concentration of Cs-137 = 4×10-7µCi/ml, I-133 = 5.92×10--4 µCi/ml, Sb-124 = 4.7×10--5 µCi/ml and maximum dose rate 0.22 µSv/hr were measured from the water surface test tube of the sipping test. The water pool conditions have been measured, i.e. conductivity 1.57 µS/cm, pH 5.58 - 7.14, temperature 26.76 ˚C, water level 6.31 - 6.43 m and air contamination area 0.1367 Bq/m3 and also they meet the criteria. For this reason, the results indicate that there are no leaks detected in the SFs.
This paper introduces three books recently published on transport and storage of radioactive materials. The transport and storage technologies have been established by accumulation of experiences and researches. Such works should be shared and used by readers and the future generations to advance the technology effectively.
The first book is “Safe and Secure Transport and Storage of Radioactive Materials” published in 2015. It reviewed best practice and emerging techniques in the following areas.
- Operational safety covering functional requirements, training, public relations and emergency response in the nuclear transport industry.
- Package design and performance for transport, highlighting mechanical and thermal considerations, radiation protection, subcriticality, and operational aspects of sea transport.
- Packaging, transport and storage of uranium concentrates and uranium hexafluoride, fresh and spent fuel, large radioactive components and medical and industrial radioactive materials.
- Long-term storage and subsequent transport of spent fuel and high-level radioactive wastes.
The second book is “Basis of Spent Nuclear Fuel Storage” published in 2015. It firstly addressed safety standards and codes for spent fuel storage. Then, demonstrative test results conforming the safety requirements were provided as follows.
- For metal casks storage, heat removal, containment, subcriticality, structural integrity, seismic performance, severe accidents, interaction between transport and storage.
- For concrete cask storage, heat removal, shielding, structural integrity, earthquake resistance, long-term integrity.
- For spent fuel, integrity in normal and accident storage conditions, and inspection method for ageing.
The third book is “Basic of Transport and Storage of Radioactive Materials” published in 2018. After reviewing the preceding books, it provided new and advanced information so that readers could have wide spectrum of the technologies. It systematically provided findings from lots of valuable researches on safety and security of transport and storage of radioactive materials under normal and accident conditions that have an impact on basis of safe regulations, designs, and operations. Characteristically, it described safety and security of transport and storage, which have been rarely addressed in the same book.
The Pakistan nuclear power generation capability is progressing rapidly. Currently five Nuclear Power Plants are in operation and two Nuclear Power Plants are under construction. Pakistan is committed towards safe, secure and sustainable management of spent fuel generated from its Nuclear Power Plants operation. The spent fuel so far generated is stored at reactor spent fuel pools under water. These spent fuel pools have limited storage capacities and are not designed to accommodate spent fuel generated from lifetime operation of Nuclear Power Plants in Pakistan. Consequently, Pakistan has decided to develop spent fuel dry storage facilities for storage of spent fuel for extended periods till ultimate decision regarding spent fuel management is taken. The paper discusses the experience of managing the spent fuel at reactor pools and development process of spent fuel dry storage facilities in Pakistan.
Due to the stringent requirements after the Fukushima accident and due to stricter requirements arising from the new design extension conditions (DEC) requirements which have been adopted into new Slovenian nuclear legislation, the Krško NPP decided to implement the safety upgrade project (SUP). SUP also envisages the safety upgrade of spent fuel storage. The NPP decided to construct a new spent fuel dry storage (SFDS) system as this is much safer and reliable as a passive system compared to the existing spent fuel pool. The new SFDS is designed to DEC conditions in accordance with the West European Nuclear Regulators Association (WENRA) requirements from 2014. Some of design basis conditions are defined even stricter by the operator. The design and construction of the new SFDS, which will meet all specified design basis conditions, are a challenge for both; the manufacturer and the operator, who will manage the SFDS. The important upgrade of the Krško NPP’s safety of spent fuel storage will be achieved with SFDS successful operation. The licensing process for SFDS started in year 2017. The design conditions for SFDS were defined by Slovenian Nuclear Safety Administration (SNSA). After the redesign of the original project, the positive opinion for the construction license was issued by SNSA in January 2019. The operation of SFDS should begin in 2021.The paper describes an outline of new DEC requirements for spent fuel dry storages, along with the example to articulate some of the Slovenian DEC requirements and how these are applied to the Krško NPP spent fuel storage.
As for the concrete cask, because its canister is a welded construction, a helium leak from the canister wasn’t considered in the past. However, during long-term storage of spent fuel, stress corrosion cracking (SCC) could occur at welded parts of the canister, so that a loss of sealing performance is concerned now. To resolve this concern, we have been developing methods for detecting the leak by using a canister surface temperature change which occurs when the gas leaks from the canister. We performed leak tests using full-scale concrete cask models in 2003, and found the phenomenon that the temperature at the lid part (TT) of the canister decreases and the temperature at the bottom (TB) of the canister increases when the gas leaks from the canister. As a result, we proposed a method for monitoring the temperature difference ΔTBT (= TB-TT) instead of the internal pressure. Recently, we proposed new detection methods using only the temperature of either the lid or the bottom of the canister in consideration of easy installation and maintenance of temperature sensors. Besides, we conducted leak tests using a 1/18-scale canister model, and succeeded in the reproduction of the phenomenon that the temperature of TT decreases and the temperature of TB increases during the leak from the canister. A mechanism of this phenomenon was verified by performing numerical analysis. We also performed leak tests by using a 1/4.5-scale cask model based on the similarity law of thermal hydraulics. In the tests, air was used for an inner gas of a canister of the model, and the heat flux of the canister surface had the same value as that of the actual canister surface. Thus, the Ra number of the model could be made to coincide with that of the actual canister. Besides, the Gr number and Bo* number were almost equal to those of the actual canister. In these tests, we generated minute leaks of the inner air, and measured the temperatures of the canister surface and outside air at the inlet. Then, we evaluated an early leak detection method based on the correlation of those temperatures.
In the United States, used fuel assemblies discharged from nuclear power plants (NPPs) have historically been stored on-site using licensed dry storage systems. The duration of storage at the NPP sites was intended to be short term (20 to 40 years) with subsequent transportation to a geological repository. Due to shutdown/decommissioning of several NPPs and delays in the implementation of the geological repository, there is a need to develop solutions to manage the storage of dry storage systems in interim storage facilities.
The centralized interim storage facility (CISF) functions as an intermediate repository/staging area for dry storage systems for an extended duration of time prior to eventual transportation to a geological repository. The operation of a CISF will lead to a significant reduction in the number of independent spent fuel storage installations (ISFSIs) nationally and enable release of space in those decommissioned reactor sites that currently only maintain their respective ISFSIs.
The licensed dry storage systems are categorized as 1) metal cask systems with fuel assemblies directly loaded into the cask which is then stored on site or 2) canister systems where the fuel is housed within a thin-walled canister which is then stored in an overpack. Regardless of type, these systems need to be designed and licensed to ensure that the necessary safety functions are maintained during long term periods of storage and subsequent transportation after storage. In addition, consideration needs to be given to potential aging deterioration of component materials that may occur during operation of the storage system at the ISFSI or in a CISF.
Managing the effects of aging of the structures, systems, and components associated with dry storage is therefore, an important aspect of the extended interim storage of used fuel. Effective aging management programs require a technical understanding of the aging degradation mechanism, inspection and assessment techniques, prevention and mitigation measures (to retard the effects of aging) and, as needed, guidance on repairs or replacements for each component. Significant research is being carried to develop expertise on the various aspects of aging management including material behavior, inspection methods, criteria and long term durability.
The CISF approach to dry storage offers a significant advantage wherein an aging management program can be effectively and uniformly implemented for a wide variety of currently licensed dry storage systems. For example, the appropriate siting of the CISF location can be made to significantly minimize the potential for environmental degradation or natural phenomena.
The paper provides additional insights into design and operation of CISF with an objective to managing effects on dry storage systems for long-term interim storage. The paper also discusses innovative solutions being developed for comprehensive aging management within interim storage facilities.
Dry Storage Systems are used as an onsite storage method for used nuclear fuel. Since no country currently has an operable repository, it will be essential for many countries around the world to extend the period of operation for these systems. Nondestructive evaluation (NDE) inspections are needed to verify continued safe operation of these dry storage systems; however, elevated temperatures, dose rates, and confined entry/exit all pose unique challenges for deployment of NDE techniques as a part of aging management requirements for these dry storage systems. Therefore, robotically-deployed in situ NDE systems and techniques have been developed to address these challenges. Field trials have been conducted to evaluate the feasibility and improve the functionality of the NDE and delivery systems in real-world environments. This paper describes development of the NDE systems, presents laboratory tests on flawed mockup specimens, and highlights field trial efforts to refine the NDE and delivery systems for deployment with dry storage system inspections.
On March 11, 2011, a tremendous earthquake of a 9.0 magnitude occurred in Japan. In the Fukushima Daiichi Nuclear Power Station, fuel assemblies were stored for the Units 1 to 6 Spent Fuel Pool, common pool and dry casks. The paper reports the lessons learned from Fukushima Daiichi Nuclear Accident for spent fuel storage.
France has chosen a recycling strategy to manage nuclear spent fuels. This strategy relies on high level waste immobilization into a borosilicate glass waste form in order to meet both interim storage and final geological disposal safety requirements. Interim storage facilities at La Hague recycling plant have been designed to last more than an hundred years (concrete structure) factoring the need for a glass waste form durability using for instance natural or forced convection principles. This period of time during interim storage contributes to the final disposal cost optimization by allowing the radioactive decay of the main contributors to thermal power dissipated in the early years after glass production.
In parallel to the thermal design of the interim storage, lots of studies have been carried out on the glass thermal and irradiation stabilities. Thermal treatment experiments consolidated by modeling show that the glassy state is expected to be stable during the interim storage period, with no crystallization induced by the glass thermal history. Moreover, the impact of the radiations (beta and alpha decays) expected in interim storage has been studied by external irradiation, glass actinide doping technique and molecular dynamic simulation. The results have demonstrated that the glassy state will not be modified during the interim storage period and the glass will fully preserved its role of conditioning material.
Now only one Dry Storage Facility of Spent Nuclear Fuel (DSFSNF) is operated in Ukraine – facility on Zaporizhska NPP. Many different thermal investigations were done for ventilated containers of DSFSNF. In this work the generalization of scientific approaches to the thermal safety assessment are carried out. The multi-stage approach to the definition of thermal state of containers' group, single container, spent fuel assemblies and fuel rods was developed. Detailed thermal profiles of spent fuel assemblies and each fuel rod inside storage container were calculated. With usage of multi-stage approach the thermal simulations of accident conditions and influence of outer factors onto thermal state of containers were carried out. On base of obtained results the classification of accidents with channels blockage was developed and the most dangerous blockages were found. Results of thermal investigations were generalized and factors, which are influence on thermal state of containers, are detected. Method of spent nuclear fuel thermal state prediction and changes to the system of thermal monitoring were proposed.
Bushehr Nuclear Power Plant, commercially operated in Sep 2013, is a Russian type VVER 1000/446; its reactor core consists of 163 fuel assemblies which have 3 up to 4 years’ fuel life time. In this pressurized water reactor, removed spent fuel assemblies are transferred into a pool near the core. This pool has been designed to store spent fuels for 9 years and after this period, this pool will reach its maximum storage capacity and spent fuel assemblies will be transferred away from the reactor. The aim of this study was to design a safe and economic approach to manage spent fuel assemblies before the pool reaches to its maximum storage capacity. The study conducted in 3 stages. In the first stage, gamma and neutron flux calculated for spent fuel assemblies have 47000 burnup, 4.2 % enrichment, 4 years fuel life time and 3 and 6 years cooling time by using Origen 2.1. Dose rate and criticality by dint of Origen output were investigated for spent fuel assemblies by means of MCNPX 2.6. In the second stage, the thickness of transportation cask including stainless steel canister and the main body of the carbon steel cask was determined. Based on limitation in casting and fabrication in Iran, twelve fuel assemblies were considered for capacity of a transportation cask. Finally, the concrete module with 36 spent fuel assemblies capacity including 3 canisters and each canister contains 12 assemblies was designed. According to results, the thickness of stainless steel canister and the carbon steel cask were determined 2 cm and 35.5 cm respectively. Dose rates on the surface and 2 meter from the surface of the canister were calculated 5.23E+05 mSv/hr and 8.22E+04 mSv/hr respectively. Neutron multiplication factor was obtained 0.31463 for cask filled by air. Also, the dose rates obtained on the surface of the cask and in 2 meter from cask surface were about 10.21 µSv/h and 4.56 µSv/h respectively. It should be noted that these derived values met transportation regulations approved by Iran Nuclear Regulatory Authority. In addition, maximum dose rates on the surface and 2 meters from the surface of the concrete module were obtained 0.011 mSv/hr and 0.192 mSv/hr for front side (door side) and roof side respectively.
The IRSN (Institut de radioprotection et de sûreté nucléaire), the French technical support organization, was asked by the parliamentary inquiry Committee on the safety and security of nuclear installations to provide a report on the concepts and safety issues regarding storage of spent fuel from nuclear power reactors.
Based on its expertise in France and on its knowledge acquired during services performed abroad, IRSN examined the concepts of wet and dry spent fuel storage existing worldwide and in France, as well as the associated safety issues.
In conclusion, IRSN emphasizes that the choice of a type of spent fuel storage must be assessed with regard to the following considerations.
The two types of spent fuel storage, wet or dry, do not completely serve the same needs, as wet storage is essential for spent fuel with high residual heat and dry storage is well suited to highly cooled fuels. In any case, these two types of storage are complementary, but the choice of one or the other largely depends on national choices in terms of spent fuel management (reprocessing or not).
The type of spent fuel (UOX, MOX…) affects the choice of the type of storage, at least for a certain period of time. Thus spent MOX fuels have a higher residual heat and this decreases less rapidly. Their cooling time before being placed in dry storage is thus much longer than for spent UOX fuels.
From the safety point of view, whatever the type of storage, the key parameter is the residual heat of the spent fuel to be stored. In this respect, wet storage, which is generally used for spent fuel with higher residual heat, requires more extensive safety provisions than dry storage where safety relies on passive systems.
IRSN also considers that a particularly important point for the safety of spent fuel management operations is the control of zirconium fuel cladding ageing, which depends on the storage temperature. On this point, wet storage offers guarantees whereas, in dry storage, the ability to directly and easily examine fuel cladding is reduced.
Thermal analyses of dry storage systems use margins in their design basis input assumptions to ensure the peak cladding temperature does not exceed an established regulatory limit. Due to these margins, a best-estimate understanding of the thermal behavior of the dry storage system is generally not available from these design licensing basis models. The development of accurate best-estimate thermal models with uncertainty quantification can lead to a more efficient use of storage, transportation, and ultimate disposal systems. A thermal modeling benchmarking project included the Electric Power Research Institute (EPRI), U.S. Nuclear Regulatory Commission, U.S. Department of Energy (DOE), U.S. National Laboratories, vendors, and utilities to assess the accuracy of best-estimate models through experimental and validation efforts. This paper describes the joint round robin aimed at further assessing the accuracy of best-estimate simulations. A total of four model submissions to this double-blind benchmark are compared with temperature measurement data acquired by the DOE/EPRI High Burnup (HBU) project. For this project, actual HBU pressurized water reactor spent fuel assemblies were stored in a bolted conductive dry storage system with thermocouples placed inside the guide tubes. The thermal analysis results were collected, and comparisons made to the benchmark measurement data. Results highlight the importance of developing a refined assessment of key uncertainty terms in modeling solutions, including decay heat calculations and internal gaps impacting the conduction of heat.
Spent nuclear fuel (SNF) possesses potential security risks; consequently, nuclear security issues have to be considered and addressed for effective management of SNF. The vulnerability of SNF to theft, for constructing a radiological dispersal device, and to the terrorist attack during storage or transport is a major concern for States having such type of nuclear materials. Hence, the national policy and strategy for SNF management should include a set of goals and objectives to ensure the secure management of SNF, and should also set out the means for achieving these objectives. Egypt recognizes that a strong and efficient legal and regulatory framework is an important element of the national nuclear security regime; therefore, it has taken several steps to enhance its legislative and institutional frameworks. Currently, the Egyptian nuclear and radiological regulatory authority (ENRRA) is developing its regulatory system regarding nuclear security that includes developing the necessary regulations against which the various nuclear facilities will be assessed, setting out the licensing process and procedures along with definition of the corresponding documentation to be submitted by the applicant, and creating an inspection program. Regulatory requirements of the physical protection (PP) of nuclear material and associated facilities are considered as a cornerstone in ensuring the security of spent fuel. These requirements should be based on the current threat assessment or design basis threat (DBT) and should include provisions which if applied correctly, will ensure adequate security of SNF during storage either at reactor site or at an interim storage facility as well as during transport. The current work provides a proposal for security requirements of SNF taking into consideration the provisions of the Convention on the Physical Protection of Nuclear Material and its amendment, International Atomic Energy Agency (IAEA) guidance, and the best practices adopted by the international community.
On March 2019, the regulatory body of Japan, NRA, issued new regulations on dual purpose cask (DPC) for dry storage of spent fuel on the site, along with the licensing process of design certification on DPC. The requirements ensure consistency with those for the interim storage facility off the site and with transport regulations.
Interim away from reactor (AFR) storage facilities may be located at the reactor site using the complete infrastructure and personnel of the nuclear power plant for their operation or at a separate site using their own infrastructure as a standalone and independent facility. With the final shutdown and decommissioning of the nuclear power plants is this infrastructure for facilities at the reactor site no longer available. It becomes necessary to install a separate infrastructure and a separate team for the facility operation to become a standalone and independent facility. This paper deals with AFR storage facilities without own infrastructure and gives an overview as a checklist on the necessary considerations to become a standalone and independent storage facility.
The nuclear policy of Romania encompasses the development and use of nuclear energy and other nuclear fuel cycle activities, as well as oversight of the development and enforcement of nuclear legislation and regulations to ensure that all nuclear activities are strictly regulated and controlled to the highest standards to ensure public health and safety.
Romania has only one nuclear power plant, Cernavoda NPP, with two units in operation. Cernavoda NPP Units 1 and 2 cover approximately 18% of Romania’s total energy production. The Government has plans to further increase nuclear generating capacity through the commissioning of Units 3 and 4 of the Cernavoda NPP.
The dry storage facility will consist of 27 seismically qualified MACSTOR 200 modules. At present 9 modules are built and in operation. The spent fuel from both operating units will be stored in the MACSTOR 200 modules for minimum 50 years. After this interval the fuel will be transfered to a future disposal facility, whose location and design is not currently decided.
The work was conducted in the context of the International Atomic Energy Agency’s (IAEA) newly initiated activity on “approaches for nuclear power costs estimation and analysis” (the “Nuclear Cost Basis”, or NCB, project). The NCB provides guidelines and resources for developing consistent cost estimates and analyses covering, basically, all areas of a country’s nuclear power programme; from nuclear infrastructure development; to reactor construction and operation; to management of radioactive waste. The paper focuses on technologically mature, widely used, spent nuclear fuel storage options and technologies. Storage of spent nuclear fuel can be made At-Reactor (AR) or Away-from-Reactor (AFR) ― at Reactor-Site (AFR-RS) or Off-Site (AFR-OS) ―. These options may involve wet (water pools) and dry storage technologies (casks, vaults, silos). For each of these technologies and options, an effort has been made to synthesize existing literature and compile a comprehensive list of key factors affecting costs. This list will be used as a basis for developing standard cost categories and cost breakdown structures for costing purposes.
The investigation is made of the out-core neutron flux and burn-up at irradiated fuel stored in TRIGA PUSPATI research reactor tank. This is required to examine whether the thermal and/or fast neutron flux can influence burn-up of the irradiated fuel stored in the same vicinity of the reactor core, the fuel rack being located 1 m above the core. MCNPX code was used to simulate fast and thermal neutron flux for the reactor operating at 750 kW. In this work, the computational model was created using MCNPX version 2.7 with the evaluated nuclear data file for thermal neutron scattering law data (ENDF7) cross-section data library and using a 10 cm 10 cm 10 cm mesh model. The results show the axial distribution for thermal neutrons occurred at energy lower than 1 10-6 MeV. Thermal neutron travels at the maximum distance of 78 cm due to thermalization by moderator. Based on the maximum distance travels by the thermal neutron, the thermal neutron does not reach the storage rack located 1 m from the core, hence there is no burn-up occurring at the irradiated fuel since burn- up can only occur in the thermal neutron region. For fast neutron, the axial distribution energy is higher than 1 106 MeV and travels more than 158 cm. The reaction time for the fast neutron is too short to result in burn-up due to its fast travel.
Long-term storage management and soundness monitoring methods of spent fuel are now receiving worldwide attention. For the metal cask, pressure monitoring between its lids is mandatory. Meanwhile, in the concrete cask, its lid is welded and high sealing property is maintained, so that a leak of helium is not monitored. However, considering long-term storage, there is concern about the loss of the sealing property due to stress corrosion cracking (SCC). To resolve this issue, we have been developing a leak detector utilizing the phenomenon that the surface temperature changes at the time of the leak. Furthermore, in order to investigate the applicability of this detector to horizontal silo storage, a leak test was conducted with a small canister model in a horizontal attitude. A basket was installed in the small canister, and 12 heater rods were installed in it. In the test, after reaching a steady state, the internal pressure was changed to 5atm, 3atm, and 1atm, and the temperature data of each position in each state was acquired. When the pressure was reduced, the canister bottom temperature and the canister side bottom temperature increased. In contrast, the temperatures of the top of the canister lid and the top of the canister side surface decreased. Therefore, it was confirmed that a highly sensitive detection method is possible with each combination using these four positions as detection points. In addition, by performing CFD analysis, the phenomena inside the canister were grasped. In this case, three-dimensional steady state CFD analysis with a polyhedral mesh using STAR-CCM+® was performed. Three kinds of internal pressure (5atm, 3atm, and 1atm) and four kinds of arrangement of the basket inside the canister were combined, which resulted in a total of 12 calculations. As a result, it was confirmed that the temperatures of the top and bottom of the canister and the top and bottom of the canister side surface vary according to the canister internal pressure as in the test results. In addition, it was also confirmed that there is significant difference in temperature distribution on the canister surface due to difference in the arrangement of the basket inside the canister.
The main objective was to validate that the natural convection refrigeration designed for the ASECQ Installations, to be built in the CNA I, will be enough to keep within a range of safe temperatures, and without risk to the integrity of the Spent Fuel Element Pods .
With the continuous development of China's nuclear power projects in recent years, spent fuel has gradually entered a long-term storage stage, in which wet storage is the main way. The supervision of spent fuel storage facilities is becoming increasingly important. As China's nuclear safety regulatory system in this area is not complete, and there are few examples to refer to, it is urgent to explore and clarify the important concerns. Based on China's nuclear safety regulatory requirements and related regulatory experience of such facilities, this paper puts forward some suggestions for reference.
The French nuclear fuel cycle includes in particular the manufacture of uranium-based fuels, the reprocessing of the spent fuels, the fabrication of MOX fuels and of ERU fuels (enriched reprocessed uranium) and the interim storage of spent fuels which are not currently reprocessed (MOX and ERU).
Regarding the used MOX and ERU fuels which are currently not being reprocessed, the strategy consists in keeping them under safe storage conditions while waiting for their future reprocessing and potential use in future generations of reactors, such as fourth-generation reactors (GEN IV).
The quantity of non-reprocessed spent fuel slightly increases every year leading to the need of extended interim storage capacities.
In this regard, the French National Plan for the Management of Radioactive Waste and Materials (PNGMDR 2016-2018) provides that, given the prospect of saturation of spent fuel storage capacities between 2025 and 2035, the french nuclear power plant operator, called the operator in the following text, shall submit to the Minister of Energy by March 31, 2017, its strategy for managing storage capacity of spent fuel from NPP and the timetable associated with the creation of new storage capacities.
In response to this point, the operator submitted the safety option file for a new spent nuclear fuel wet storage facility, so-called the centralized interim storage pool.
As requested by the French nuclear safety authority (ASN), IRSN reviewed these safety options. The IRSN assessment focused on the safety approach at the preliminary design stage and the structuring design choices such as:
• the civil engineering structures, including storage pools (number, dimensions, storage capacity, subdivision) and the building shell,
• the methods of unloading, loading and handling of spent fuel assembly transport casks as well as the methods to store the spent fuel assemblies,
• the consideration of scenarios of loss of "support functions" identified by the operator, in particular the duration of autonomy retained in the event of a prolonged loss of electricity or cooling function.
• more generally, the accident situations to be taken in the design of the installation
Special attention was paid on the one-hundred-year operational lifetime of this facility with regard to the surveillance of spent fuel assemblies, the control of the facility ageing, in particular through maintenance and monitoring, and the inspectability and possible replacement of systems, structures and components important for safety
Fuel is periodically replaced in nuclear power plants (NPPs). Irradiated or Spent Nuclear Fuel (SNF, where SNF could be used nuclear fuel if reprocessing facilities are available) cools in suitable facilities, where the type and the length of time depend on plans for the ultimate disposition of the SNF, for example, reprocessing or permanent long-term storage (“extended” implies storage longer than 50 years). The paper attempts to calculate the relationships between the costs and the sizes of on-site wet and on-site/off-site dry storage facilities. This is done by estimating reduced-form equations based on publicly available data, which can be modified with more recent, detailed, or proprietary data to update or extend the analysis: the values reported here should not be considered as the only possible outcomes; they are used here to understand relative NPP SNF owner economic incentives. The paper finds that once the NPP has been decommissioned, and only the on-site dry storage remains, there might not be a cost reason (from the point of view of the NPP owner/operator) to move the SNF to consolidated facilities. However, there is a consensus that consolidated facilities (a) would be more safe and secure than dispersed on-site storage locations, (b) would facilitate final disposal, and (c) can reduce the risks perceived by local communities near SNF storage facilities.
In Switzerland, the spent nuclear fuel assemblies arising from the operation of the five NPPs are currently stored in pools at the NPP sites and, after a cooling period, are transferred to transport/storage casks which are then transported and stored in centralized dry interim storage facilities. The National Cooperative for the Disposal of Radioactive Waste (Nagra) has proposed deep geological disposal as the solution for the management of all radioactive waste. Pre-disposal activities, in particular for the spent fuel encapsulation facility and related unloading/loading and handling operations from the transport/storage casks into the final disposal canisters, are safety-relevant operations. Nagra therefore initiated several studies and RD&D activities aimed at assessing spent fuel mechanical performance, but also at developing concepts for handling of consequence scenarios. Concerning the RD&D program, the main objective of the investigations is to assess the response of spent fuel rods to mechanical stresses corresponding to normal conditions and accident scenarios by means of experiments on PWR spent fuel rod segments. The experimental campaign is conducted at JRC Karlsruhe, with the focus on the effect of hydrogen load, hydride distribution and pellet/cladding interaction on the cladding integrity. Other studies are currently under development to investigate the deterioration of the cladding properties resulting from Delayed Hydride Cracking (with Paul Scherrer Institute), as well as the deterioration of the FA structural material for long-term dry storage conditions (with Framatome GmbH). Furthermore, a conceptual study is under development to establish specific technical requirements for the encapsulation facility, focusing on fuel handling, retrieval and packaging operations. The main scope is to ensure the safe management of any damaged and degraded fuel and to implement measures for the mitigation of accident scenarios. Key aspects and main achievements of these ongoing programs are presented here.
Two approaches usually are used at thermal safety assessment of the spent nuclear fuel storage: conservative calculation and detailed simulation of thermal state. For the second one the multistage approach was developed which allows to calculate thermal contours of containers' group, single container, spent fuel assembly and fuel rod with taking into account their mutual influence and influence of outer factors (weather conditions and solar irradiation). All calculations were done for the Dry Spent Nuclear Fuel Storage Facility on Zaporizhska NPP. Temperature state of containers on the open-air storage platform under wind influence were received. With taking into account the changes inside container's ventilation system the temperature of spent fuel assemblies (Fig.1) and maximum temperatures of each fuel rods were calculated. Detailed data about thermal state of spent fuel during dry storage are planned to be used at thermal stress calculations and ageing assessment.
Research is carried out under CRP-20605 (SPAR-IV).
Severe accidents at the Fukushima Dai-ichi nuclear power station became an important lesson learned to know the water temperature distribution at spent fuel storage pool (SFSP) of G. A Siwabbessy nuclear research reactor in Indonesia. When the active cooling system was not functioning properly, the knowledge of the cooling water temperature in the SFSP became an important parameter related to SFSP safety. The research objectives to determine the cooling water temperature distribution in G. A. Siwabessy SFSP when the active cooling system was failure in function. The experimental method is used to know the temperature distribution in the pool water when the active cooling system is turned off for 87.18 hours. The experiment results show that the highest temperature of pool water when the active cooling system failure was 26.89°C. With the present spent fuel, the results obtained show that the temperature of the water in SFSP does not exceed the temperature value which can cause high evaporation of water and does not cause danger to the overall spent fuel integrity.
During reactor operation, the mechanical properties of a nuclear fuel rod are radically altered. After discharge, alpha-decays and accumulated radiation damage or other processes associated to potential thermal variations occurring during interim storage, contribute to further ageing of the spent nuclear fuel (SNF). However, during all stages of the SNF management (han-dling, retrieval, packing and transportation to final disposal or reprocessing) safety must be guaranteed.
Assessment of the SNF mechanical stability against external stresses, which might be accidentally applied, requires rep-resentative reference data. Explicit tests simulating accidental conditions are conducted in the hot cells of JRC - Karlsruhe, in the frame of a multi task collaborative research programme. Three-point bending and gravitational impact devices were developed and installed in the hot cells to investigate the SNF rods response under quasi-static or dynamic loads. Load-deflection curves are generated in the 3-point bending tests, whereas a high-speed camera records the rod rupture during impact tests.
Results from investigations on LWR commercial fuel rods over an extended burn-up range are presented in this paper. SNF segments, pressurized at the original fuel rod pressure after discharge, were subjected to bending and impact tests. Similar masses, significantly less than a single fuel pellet, of fuel disperse upon pin rupture in both types of experiments. Only fuel fragments from the immediate vicinity of the rod fracture release. An image analysis methodology was developed to elaborate the sample’s behaviour under dynamic loads. Optical and electron microscopy were used to observe the morphology, orienta-tion and population of the cladding hydrides, whereas the overall hydrogen concentration in the cladding was measured with hot extraction technique. Size distribution analysis on the released fuel particles was also performed. The study is augmented by modelling approach to evaluate the individual phenomena and parameters affecting the SNF properties.
Accumulation of spent nuclear fuel in PWR power reactor pools is facing saturation limit within 5 years in South Korea. Though the national policy will be discussed again through public hearing process, it seems very clear to imply dry storage technique for the first management step out of reactor pools like other countries. Spent nuclear fuel (SNF) integrity evaluation R&D work has been performed for lower burnup (less than 45 GWd/tU) range for 5 years in order to produce initial SNF characteristic properties and anticipate aging effect during dry condition for several decades. This project produced non-destructive examination data which are essential prior to the destructive testing. The former data set includes visual exam, defect scan, dimension measurement and gamma scan. The later data set includes creep testing, hydride reorientation testing, delayed hydride cracking, ring compression testing, optical microscope analysis, hydrogen contents analysis and mechanical properties testing. This project also tried to evaluate fuel assembly hardware integrity including spacer grid, welding points between components, and bulge joint for real SNF components and for unirradiated components by charging hydrogen to simulate SNF. In order to anticipate SNF degradation for several decades, modeling of each single effect like creep and hydride reorientation have been done and comprehensively merged into a newly developed SNF performance platform which deals with thermal profile among SNF rods. Based on achievement for lower burnup range, SNF R&D infra could be expanded to high burnup range successively.
Two potential cladding degradation mechanisms have been the focus of regulatory authorities’ reviews when evaluating applications for storage and transportation of spent nuclear fuel under dry, inert atmosphere conditions: thermal creep and hydride re-orientation. A review of the thermal creep mechanisms in the low and high stress regions and their dependence on applied stress, as supported by experimental work, leads to the conclusion that creep failure is highly unlikely for internally pressurized spent fuel rods. Through a process of elimination, the formation of radial hydrides can be assessed to be minor or completely eliminated for several types of BWR and PWR claddings; claddings that would benefit from additional investigations include RX and pRX PWR claddings when hydride dissolution upon heating occurs in a temperature range sufficiently high for complete hydride dissolution, but too low for any significant radiation damage annealing.
The first German package design approval certificates for a dual purpose casks intended for loading with damaged spent nuclear fuel were issued recently. BAM as part of the competent authority system in Germany carried out comprehensive assessment procedures with respect to the mechanical and thermal design, the release of radioactive material and the quality assurance aspects of manufacturing and operation. Packages for the transport and storage of radioactive material have been assessed by BAM for many years, thus the common assessment procedure is well-known and good practice. Up to now only SNF without defects or HLW with well-defined properties were designated for long-term interim storage and transports afterwards. Due to Germany’s nuclear phase out all other kinds of spent nuclear fuel in particular damaged spent nuclear fuel shall be packed now. Damaged spent nuclear fuel needs a tight closure with special encapsulations and clearly defined properties in Germany. In addition, these encapsulations shall be long-term durable, because they are not accessible after loading in a packaging within periodical inspections. The main difference to standard package components is that encapsulations with a permanent closure achieve their specified conditions not after manufacturing but only during operation, after loading and closing. To ensure compliance with the specific conditions, special measures for quality assurance are necessary during operation of each encapsulation, e.g. drying and sealing, which were assessed by BAM. The present paper gives an overview of the conducted assessment from BAM and point out the findings concerning to the special closure lid of the approved encapsulation, which is screwed and welded. A wide verification concept was necessary to show the specific tightness under transport conditions. Together with quality assurance measures during first operation steps these encapsulations with damaged spent nuclear fuel can then be handled like standard fuel assemblies in approved package designs.
Spent Fuel management in Spain is open cycle. Nowadays the Spanish NPP (7 PWR and 2 BWR) have constructed or designed ISFSIs (Independent Spent Fuel Storage Installation), however a Centralized Storage Installation will be operating in the future. The first Spanish ISFSI was installed in Jose Cabrera NPP (14x14 PWR Fuel) due to the definitively cease of the NPP operation in 2006. At that time Enusa Industrias Avanzadas developed, and made available to its customers, its resources and capabilities applicable to all the phases of the management of the irradiated fuel. These capabilities and resources are especially focused on two major areas: engineering and on site fuel services. These paper will be focused on engineering developments in regard to Spent Fuel Management.
In engineering area, a Methodology for the classification (damage/ no damage) of fuel assemblies has been developed, considering characteristics as internal pressure, hoop stress, hydride lens in spalled oxide positions, stress corrosion cracking in the top nozzle-skeleton sleeves, etc. A detailed data base for each NPP has been designed and developed taking into account the most important fuel assembly characteristics. Different type of inspections (visual for integrity, In- can sipping and UT for fuel rod leaks detection, visual for oxide spalling…) are performed to complete every fuel assembly characterization before classification.
With the objective of decrease the number of fuel assemblies classified as damaged, some developments have been performed: (1) design, licensing, manufacturing and installation of a device named ESPIGA to solve the handling problems for oldest fuel assemblies affected by intergranular stress corrosion cracking on top nozzle sleeves; (2) specific analysis to detect fuel assembly leaks; (3) methodology to assure the fuel rod integrity during transport considering hydride lens deep up to 40% of cladding width ; (4) methodology to calculate cladding hoop stress to assure the fulfilment of the regulation limit (90 MPa in Spain for low burnup fuel assemblies if storage temperature is above 400ºC).
Spent fuels (SF) assemblies from Paks Nuclear Power Plant (Paks NPP, Hungary) are placed in Spent Fuel Interim Storage Facility (SFISF) since 1997. The SFISF is a modular vault dry storage (MVDS) type design accommodating SF after a minimum of a few years of cooling time in the reactor decay pool. The SFs are stored individually and separately in the vault modules (VM) in airtight sealed fuel storage tubes (FST) filled with inert gas. Decay heat rejection is achieved by buoyancy driven air flow through the vault, passing over the exterior of the array of storage tubes.
The capacity of the SFISF was planned on the total amount of the SFs arising from the planned 30-year lifetime of Paks NPP. To store these SFs a 33-vault facility was designed with 450 FST in each vault. Until now all together 24 vaults have been constructed.
Sixteen vaults were built with 450 FST in each vault. To make the storage economically more efficient the number of FSTs was increased from 450 to 527 in the last eight vaults. This was provided by use of the built-in reserves of the design and the development of analyses techniques making it possible to reduce the conservatism in calculations. According to this modification the total capacity of the SFISF was increased by around 9%.
At the millennium a decision was made to extend the lifetime of the Paks NPP with addition 20 years, resulting a significant growth in the amount of the SFs. In order to adjust the storage capacity a review of the design was carried out. The structural analysis showed that a number of 703 FSTs could be installed into the same geometry by modifying the charge face structure (CFS). Based on this number the total capacity could be increased by almost 20% compared to the original design.
Considering the initial few years of cooling period and applying it for the whole storage facility the heat load could be higher than the design criteria. However, with the rearrangement of the SFs cooled for many years in the FSTs it is possible to solve this issue. The decay heat production of SFs stored for many years decreased to a level at which it is possible for them to be placed in a higher density redesigned vault with the new CFS design. By transferring the older SFs to the higher density vaults there will be enough free positions to place the newer SFs arriving from the NPP. Construction license with the newly increased storage arrangement was issued by the nuclear authority in 2017.
The paper describes the design, modelling and licensing process of this capacity enhancement.
Over 35 000 tons of used fuel have been processed in La Hague so far, and more than 8000 MOX and 7200 ERU Fuel Assemblies have been produced from recycled material. These experiences together with fairly stable yearly production levels for 40 years clearly demonstrate the industrial maturity of those technologies.
In France, used fuel reprocessing and recycling facilities continue their operation while enhancing the range of LWR and RR used fuels to be treated and performing investments to both increase competitiveness and secure long-term operations. This leads to continuously develop and implement new technologies, those resulting from constant interactions between R&D teams, mainly from the CEA, engineering department and operating facility teams.
These improvements include since 2010 the implementation of a Cold Crucible Induction Melter (CCIM) in an existing and very highly active facility (R7) at La Hague. In comparison with the Hot Melter, the Cold Crucible Induction Melter allows to operate with a higher throughput and a higher elaboration temperature. This technology makes possible the conditioning of two different types of effluents (from rinsing operation and with corrosive solution) with different glass formulations. Additional recent examples range from the first silicide fuel reprocessing campaign completed in 2017 at La Hague thanks to R&D developments enabling the process qualification, the deployment of a prototype innovative technology in order to provide a 360 degree weld with no physical nor visual access, and the completion of the dismantling of one of the former fission products’ evaporator using laser technology.
R&D developments are continuously being performed to enlarge the range of used fuel capabilities to meet customer/market needs as well as to improve operating standards and secure long term operation
Development of atomic power engineering on a global scale has made it necessary to address problems associated with spent nuclear fuel (SNF) management. Operation of nuclear facilities resulted in accumulation of large amount of SNF with various compositions and geometries. The SNF is accumulated both during electric power generation at NPPs and operation of naval propulsion reactors by surface and submarine fleet, and during research and development of new approaches to fuel management carried out at nuclear research centers.
There are two competing approaches to management of generated SNF. One approach is based on long-term storage with subsequent direct disposal (open cycle), while the other one is connected with radiochemical reprocessing (closed cycle). The Russian Federation adopted a strategy of closed nuclear cycle with SNF reprocessing and recycling of recovered products resulted from reprocessing. Implementation of this strategy is closely connected with SNF reprocessing carried out by Mayak PA.
In the geological disposal of high-level radioactive waste, the amount and properties of waste buried deep underground depends on various process conditions, starting with nuclear fuel. To reduce the environmental load of geological disposal, we evaluated the effects on vitrified waste heat generation of fuel burn-up, spent fuel cooling period, separation of heat-generating nuclides, and waste loading of vitrified waste, which are closely related to the design of the repository. The timing of the appearance of the maximum temperature of the buffer material in the repository depends on the heat-generating nuclides, including short-lived nuclides Cs and Sr and their daughter nuclides and long-lived minor actinides contained in vitrified waste. The generation and accumulation of these nuclides is related to the fuel burn-up and the cooling period of the spent fuel. We examined the effect of the minor actinide separation ratio, which is related to the spent fuel cooling period, on the waste-occupied area in repository to reduce the amount of vitrified waste with high loading, assuming that molybdenum and platinum group metals were separated. Based on these results, the idea of cross-sectoral and integrated research on spent fuel, nuclide separation, vitrified waste, and geological disposal was examined to present technical options that contribute to load reduction in geological disposal.
The paper describes the opportunity of reprocessing damaged multi design fuel assemblies (of PWR, BWR or VVER origin), which is an effective solution in terms of global risk reduction. The paper also broaches the necessary associated packaging and transport logistics, the legal framework, and the management options for separated materials.
A number of European reactors (PWR, VVER1000, BWR) use fuel assemblies from different fuel designers and manufacturers. When new fuel designs are proposed or new manufacturing processes or sites qualified it may happen that a higher number of fuel defects occur before reaching design robustness and maturity. Some earlier fuel designs do not allow repair, i.e. fuel assembly disassembling and reassembling in order to extract a damaged fuel rod, inspect it and repair and replace it with a dummy rod. In other cases rigid fuel designs and the fuel assemblies’ condition after a few cycles of irradiation may introduce a high risk of fuel rod rupture during its extraction. All this together with the restricted accessibility of the reactor pond during normal reactor operation make it very difficult to inspect fuel defects and determine their root cause. Consequently defective fuel has to be classified in the worst damaged fuel category (as compared to “only” gas leaking fuel) along with fuel debris which is almost always present. In addition, storing such damaged fuel assemblies in the reactor pond may lead to operational and safety challenges for the reactor operator. The defective fuel management is therefore a matter of concern for all nuclear operators.
Reprocessing such damaged fuel assemblies allows to get rid of the above mentioned risks and challenges. This way defective fuel is eventually replaced by just a few final vitrified and compacted residue packages, specifically designed for long-term storage pending transport to future geological disposal facility.
A wide portfolio of industrial and appropriate solutions adaptable to several types of defective fuel assembly including conditioning solution, transport cask designs compatible with any fuel designs to be reprocessed are already available, such as Quivers/Fuel rod canisters, Fuel Rod Capsules and Capsule canisters for wet or dry storage or transport. Various cask designs exist for that purpose and specific operations are implemented for preparing the defective fuel assemblies for transport.
Over several decades Orano has successfully performed the treatment of various type of defective coming from all around the world. In order to transport the defective used fuel assemblies/rods to the reprocessing plant, Orano proposes a comprehensive range of qualified and proven solutions.
More than 460 leaking and damaged fuel assemblies as well as capsule canisters and quivers have been reprocessed at La Hague, using normal process and tools with specific operating conditions and monitoring. Orano pursues developing technologies and dedicated tests in operational conditions to enlarge its portfolio to reprocess new design of defective fuel to meet its customer expectations.
In addition to many shipments of defective fuel assemblies performed for EDF needs, Orano has, also, a strong international experience in transportation of defective fuel assemblies: shipments were performed from Germany, Switzerland, Belgium, Italy etc. to the La Hague plant in France with a large range of TN casks family.
Reprocessing is the best way to fully manage defective fuel in a risk mitigating approach.
Orano continues developing technologies that meet customers’ needs satisfying stringent safety requirements.
The experience of nuclear fuel cycle facilities operation shows that explosions during reprocessing of radioactive material could lead to release of radioactive elements with consequences for environmental. The root cause of many of them is chemical interactions with heat and gas generation. Hazard identification methods have been developing for many years, but the specifics of the nuclear industry requires to adapt the approaches. The chemical processes are going to be used for solving the radioactive waste problems that means that the agreed by the scientific community approaches of the safety assessment should be developed.
U and Pu in spent nuclear fuel are obtained as separate and pure streams by the standard reprocessing method, Purex solvent extraction. Economical and technological conditions and safeguard concerns have not allowed the standard Purex (and recycling) to be an integral part of the nuclear fuel cycle. Easier and cheaper methods to recover U and Pu in spent fuel are worth taking into account. Complete co-processing, based on a Purex scheme, is probably the easiest way of separating U and Pu from spent fuel; in addition, it is advantageous with regard to safeguards because it does not yield pure Pu. Products of complete co-processing are two separate mixtures: (1) U+Pu and (2) FPs+MAs (Fission Products and Minor Actinides). Because U+Pu mixture obtained from spent LWR fuels has a total fissile content of roughly 1.5 weight percent, which is not suitable for LWRs, special approaches are required for recycling in case of complete co-processing. The other product is High Level Waste by definition, which does not contain any isotope usable for energy generation.
This study focuses on a scenario in which, after a proper precooling period, the complete co-processing is applied, and U+Pu and FPs+MAs are obtained. Then, these mixtures are stored: U+Pu as a potential reserve to be utilized in energy production and FPs+MAs as waste to be permanently disposed of in the middle and/or long run. The purpose is to investigate effects of such a scenario on the back-end of the nuclear fuel cycle. Radiotoxicity levels and compositions of U+Pu and FPs+MAs products of the complete co-processing are determined as a function of length of the storage period, and several options regarding what to do with them in the long term is discussed and compared.
The paper contents one of the approach to spent fuel (SF) recycling with multi-recycling entire amount of the U and Pu after the reprocessing. The description of the main concepts of U-Pu fuel which developing in Russia is given. The status and future plans for REMIX-concept program development is described in the paper.
A new design for the PWR fuel assembly has been proposed, in a previous work, for direct use of the PWR spent fuel without chemical or dry processing has been proposed. The proposed designed assembly consists of four zircaloy-4 tubes. Each tube contains 7 or 8 CANDU fuel bundles stacked end to end. The zircaloy-4 tube has the same inner diameter of CANDU pressure tube. The PWR spent fuel bundles will be transferred directly to CANDU reactors without processing. The CANDU reactor is preferably be built in the same site to avoid the problem of transportations. In the current work, a different case has been studied for improving the uranium utilization and for reducing the high level waste. Generally, the calculations resulted in that the burnup would be increased by about 35%. The proposed strategy would reduce the high level waste. Moreover, direct recycling of the spent fuel would degrade the plutonium vector which enhances the proliferation resistance.
Management of severely damaged spent fuel and corium plays an important role in the nuclear safety issues for the nuclear power plants. The calculation of burned fuel inventory is required for determining the composition, activity of core melt and in the estimation of the radiological source term in the environment. Isotopic inventory of the burned fuel at the time of the accident of Fukushima Daiichi Unit 1 (FD-U1) was calculated using Monte Carlo analysis MCNPX 2.7 code linked to depletion calculation code CINDER'90 and ENDF/B-VII.0 cross section data library. The reactor core model results were validated with experimental measurements which was carried out by Japan Nuclear Energy Safety Organization (JNES) and verified with published results using ORIGEN-Code by Japan Atomic Energy Agency (JAEA). The verification comparison was in good agreement for all the radionuclides, and more radionuclides were obtained using MCNPX-Code. The total activity of the burned fuel at the time of the accident was 9.86E+19 Bq and after 50 Yrs. was 1.89E+17 Bq and the higher inventory concentration in the fuel was dominated by the trans-uranic elements. Also, the specific activity was calculated for the inventory at the time of the accident and after 50 Yrs. And found to be 1.84E+15 Bq/gm and 5.86E+12 respectively.
Several countries reuse MOX fuel (reactor-grade (RG) and weapons-grade (WG) Pu) in the thermal reactors, but within no more than 30% of the total core loading. Many years of operational experience with MOX-fuelled cores along with well-developed technologies in the management of MOX fuel demonstrate possibilities in extending MOX fuel share in commercial nuclear power plants. Therefore, this paper deals increasing weapons-grade plutonium disposition rate (mPu) in the thermal reactors using differenced methods as: reducing the burnup, reducing the residence time of the MOX assemblies in the reactor, increasing the fraction of MOX assemblies in the core, and reducing the plutonium enrichment in the MOX fuel. The results showed at EOC for 100% MOX fuel that the mPu were: 1320 and 930 kg/year and the 240Pu fractions were: 31 and 45% for VVER and RBMK respectively. The mPu was increased: 930, 985, 1076 and 1590 when the plutonium enrichment was reduced: 1.8, 1.6, 1.4, 1.0 % respectively. The mPu was increased by 45% when using four UO2 fuel cycles and one MOX cycle instead of three UO2 fuel cycles and one MOX cycle.
Recycling operations have been mastered for long in France, from the plutonium separation to the irradiation of MOX fuel, as France committed itself towards recycling plutonium in PWRs since 1987. Today, the French reactors using MOX are operated according to fuel management allowing equivalent performance of energy supplied with the same reliability as those using UO2 fuels.
The paper first presents the experience feedback obtained up to 65 GWd/tHM (rod average). Fuel microstructural evolutions under operations as well as the behavior of fission products have been thoroughly examined. A somewhat higher fission gas release is observed compared to UO2 fuel mainly due to the higher power levels of the MOX fuel and its more heterogeneous microstructure. To keep the parity with UO2 in the future, MOX evolution based on advanced microstructures is considered to provide the required performance. In that respect, the CHROMOX microstructure obtained by Cr2O3 doping shows an enhanced homogeneity notably with smaller primary blend agglomerates and increased matrix grain size. With these evolutions, internal pressure margins are anticipated and better retention of gaseous fission products in accidental conditions by reduction of restructured areas.
To sustain the use of MOX fuel in the future, the second part of the paper presents the adaptations to be implemented at the MELOX production plant to face the inherent degradation of the Pu isotopic vector of MOX fuel and its higher Pu content from increased core management cycle length.
In addition, Pu multi-recycling strategies in LWRs are studied with new fuel technologies. In order to be able to use low quality Pu in a PWR spectrum, fissile uranium needs to be added. With the CORAIL-A option, developed by Framatome and Orano, the assembly contains about half of MOX fuel rods and the remaining as UO2 rods. By contrast, the MIX fuel assembly contains only MOX rods with an enriched uranium matrix that compensates the Pu degradation. Development of those fuel technologies, that could be coupled with the most advanced Framatome fuel assembly design GAIA, will offer flexibility to switch to future technically and economically robust advanced cycles in current or future LWRs with a limited impact to the reactor design and its performance. These developments will allow implementing efficient solutions bridging the gap with the potential development of GEN IV reactors.
Various ways of using reprocessed uranium and plutonium in closed nuclear fuel cycle (NFC) of thermal-neutron reactors by reprocessing spent nuclear fuel (SNF) from nuclear power plants with release of these materials and manufacturing of secondary fuel are described.
In the scope of Sodium Fast Reactor (SFR) spent fuels recycling, the chemical composition and irradiation conditions require a specially-adapted head-end treatment process to quantitatively dissolve the plutonium. Dissolution studies are therefore being performed at the CEA ATALANTE facility in Marcoule research center. The R&D is based on the use of experimental irradiated Phenix fuels. The dissolution studies include a primary dissolution step in nitric acid medium. Solid residues are then separated by filtration and dissolved applying an oxidizing digestion process which is more efficient for dissolving high Pu-content particles and metallic elements. The composition of SFR spent fuel residues and their masses are presented and linked to the initial fuel compositions and irradiation conditions. Plutonium recovery rates along with the dissolution process, including primary dissolution and oxidizing digestion, are presented. The efficiency of the oxidizing digestion of dissolution residues is also discussed, illustrating the advantages of this chemical step not only for plutonium recovery but also for metallic element dissolution.
In accordance with the French Act of 28 june 2006 on the sustainable management of radioactive materials and waste, the CEA in partnership with EDF, Orano and Framatome, has studied prospective scenarios using different fuel cycle options: open cycle, recycling of plutonium and uranium in PWRs (current option for the French nuclear power fleet), multiple recycling of plutonium in SFRs and multiple recycling of plutonium in PWRs.
This information has been submitted by the CEA to the Ministry of Energy within the scope of Article 51 of the Ministerial Order dated 23 February on the French National Radioactive Materials and Waste Management Plan (PNGMDR).
Rather than suddenly switching over to the large-scale deployment of fast reactors as assumed in past scenarios, it is now deemed preferable to ensure the progressive implementation of this technology through successive phases: each phase involves the more significant deployment of fast reactors with its own growth objective.
Phase A corresponds to the current state of the French nuclear reactor fleet wherein plutonium and uranium are recycled in mixed-oxide (MOX) and enriched reprocessed uranium (ERU) fuels in pressurised water reactors (PWR).
Phase B consists in recycling spent MOX fuel from PWRs in a limited number of SFRs. The objective of this phase is to stabilise the quantities of spent MOX fuels from light water reactors.
Phase C is designed to be able to stabilise the plutonium inventory by deploying a symbiotic fleet comprising UOX-PWRs, MOX-PWRs and SFRs.
The objective of phase D is to deploy a fleet of reactors that no longer burns natural uranium. There are two possible options for a nuclear fleet that can generally be considered as self-sufficient, i.e. D1, a homogeneous fleet with 100% SFRs, and D2, a mixed fleet comprising breeder SFRs producing plutonium and PWRs fuelled with 100% MOX to burn this plutonium.
A phase 0 was also defined. This phase corresponds to a hypothetical French fleet having operated in an open-cycle configuration only.
However, SFRs may not become economically competitive in the next few decades if uranium resources remain readily available, and MOX spent fuels may start to pile up at the back-end of the fuel cycle unless alternative plutonium management solutions in PWRs are found. In this study, advanced fuel batches, called CORAIL and MIX, are applied to enable multiple recycling in standard PWRs. The main objectives of these scenarios consist in fast recycling of all spent MOX fuels and in stabilizing the plutonium inventory as well as all spent fuel stocks.
For this first study, we consider CORAIL assemblies composed of 181 UOX rods and 84 MOX rods (a mixture of plutonium and depleted uranium). This configuration was studied in the early 2000s. As the enrichment of UOX rods is maximised at 5%, the plutonium content is adapted to make up for its loss of fissile quality with each recycling phase.
MIX assemblies consist of identical rods filled with plutonium and enriched uranium to a content in U235 suited to compensate for the isotopic degradation of plutonium. Three plutonium contents have been considered in our study : 8%, 9.54% and 12%.
The intial conditions of the scenarios corrsponds to the current French fleet with its 58 PWR units generating around 420TWhe per year. This annual electricity production was taken as being constant over time. A future reactor lifespan (PWRs and FRs) of 60 years is considered. A lifespan of 50 years was assumed for the fuel cycle plants (reprocessing and manufacturing).
For scenarios involving the progressive deployment of SFRs, the start-up of commercial SFRs is expected occur 25 years after the industrial commissioning of the Astrid-600 MW reactor (assumed in this study to be in 2039), i.e. in the mid-2060s. This timescale takes into account the need for sufficient feedback from the operation of Astrid and for realistic lead times for technical and regulatory actions.
For scenarios involving the multiple recycling of Pu in PWRs, it has been considered that the industrial deployment of CORAIL and MIX concepts would be theoretically possible in 2045, a timescale that seems at this stage in the studies to be reasonable for qualification of these new fuel products.
According to a series of interstate agreements of the Russian Federation, spent fuel assemblies from the Russian-origin reactors are subject to return to the Russian Federation for interim technological storage and subsequent reprocessing. Meanwhile according to the legislation of the Russian Federation, products of SNF reprocessing are subject to return to the Supplier’s state. The principle of activity equivalent for the imported SNF and the reprocessing products returned to the Supplier’s state is used in the Russian Federation to determine the volume of reprocessing products to be returned, taking into account the natural decay of radionuclides for the period of technological storage. However, there is no uniform approach to determine the activity equivalent criteria. The paper describes the main approaches implemented in the Russian Federation to determine this criteria and shows prospective ways to its definition.
Various issues of deep geological disposal of radioactive waste, including alternatives to geological disposal, multinational approaches and costing / financing aspects are being discussed in the public and political sphere
in a recurrent manner. EDRAM believes, as a group of senior executives from national agencies for implementing radioactive waste disposal in their respective countries, that drawing on international expertise, experience and collaboration is of great value and leads to better solutions for the safe implementation of radioactive waste disposal. Based on this belief EDRAM discusses strategic issues and technical and management matters, with a view to benchmarking and establishing best practices, develops a common understanding of waste management issues among implementers and positions thereof and coordinates actions in relation to international organisations. EDRAM continually exchanges information on these matters within the group and with international organizations and understands differences and commonalities among them deeply in order to be able to explain them to its stakeholders. In the paper summarized are some of major recent outputs from this discussion.
All options for generating power from nuclear energy generate radioactive waste products that will require permanent isolation from the biosphere. Choices made regarding nuclear fuel cycle options, including decisions for recovery and re-use of fissile material from irradiated fuel, have the potential to affect the waste stream characteristics such as mass, volume, radioactivity, and thermal power, but no options eliminate the need for robust isolation of wastes. Decades of experience has produced an international consensus that deep geological disposal is the preferred method for achieving permanent disposal. The paper reviews published results of safety assessments for deep geologic disposal concepts that have been proposed in the United States, Sweden, France, Switzerland, and other nations to provide insight into the waste form aspects that most affect the long-term performance of repository systems. Disposal concepts considered include geologic repositories in multiple rock types in both saturated and unsaturated environments. Additionally this work evaluates how repository performance may be affected by hypothetical waste form modifications from changes in fuel cycle choices.
Andra is preparing an application for the creation of a geological disposal facility in East of France, Meuse Haute-Marne area: Cigéo Project. The project will be implemented in the vicinity of Andra’s Meuse Haute Marne Underground research laboratory which has been operated since 2000. The project aims at disposing of vitrified HLW produced by the reprocessing of spent fuel, as well as a range of ILW, including metallic parts of the fuel assemblies separated during reprocessing and various operational wastes. Cigéo design capacity covers existing HLW and ILW in France as well as wastes which will be produced in the next decades by the operation and decommissioning of all existing French nuclear facilities. ILW will be disposed of from early stages of the operational phase while HLW will be accommodated after a few decade storage phase for thermal decrease. A total 100 year operational period is envisaged. By law Cigéo project is designed consistently with reversibility requirements. This includes a number of means to make it possible to modify decisions along the process, such as progressivity and flexibility in the implementation of disposal cells. Within this framework the very first period of the project will be a pilot industrial phase. A particular emphasis is being given to the « adaptability » of the project, which is considered as part of its reversibility and is expected to be assessed within the application. Provisions are made so that next generations can adapt the operational process to accommodate potential changes in the French spent fuel and waste management policy and strategy. This includes the potential direct disposal of spent fuel.
The amounts of waste generated in the nuclear power lifecycle is small compared to other power generation options and normalised to power produced. In particular, because of the enormous energy density in uranium, nuclear power plants produce much smaller quantities of waste than fossil plants. Although there are several back–end management options that result in different waste forms for countries generating spent fuel and high-level radioactive waste, a geologic disposal capability is required.
High-level Radioactive Waste (HLW) and/or Spent Fuel (SF) need technologically advanced treatment and management procedures from interim storage to final disposal. To prevent any negative impact on the environment or and human health, HLW and SF must be adequately isolated. Disposal in a Deep Geological Repository (DGR) is internationally recognised as the most technologically developed and safest approach to isolating these wastes from the biosphere. Development of a DGR involves high fixed costs that carry an associated economy of scale. A DGR with a capacity of 10,000 metric tonnes can cost little more than one to dispose of 5,000. This means that smaller nuclear programs could benefit greatly from the opportunity to participate in a Multinational Repository (MNR).
The MNR concept provides a shared solution to the challenges of SF and HLW disposal. The concept involves a service provider country developing a geologic repository and accepting SF from several customer countries. Although financing is an issue shared by all repository projects, a MNR project presents a unique case regarding issues associated with the sources of funds, timing of revenues and expenditures, and risk allocation. Different international organisations are approaching this issue from diverse aspects. Recent developments regarding the identification of financing approaches for an MNR have been observed among different fora and will be presented in the paper. These activities include actions of different intergovernmental and international organisations (i.e. IAEA, OECD, WNA), however this paper will focus mainly on results of recent work done by the International Framework for Nuclear Energy Cooperation’s (IFNEC) Reliable Nuclear Fuel Services Working Group.
Finland and Sweden are the countries with the most advanced plans for final disposal of spent nuclear fuel (SNF). The OECD Nuclear Energy Agency's Forum on Stakeholder Confidence (FSC) has evoked both countries as good examples in the use of a 'partnering' approach, designed to achieve both a licensable site supported by the community and a balance between fair representation and competent participation. While both countries are consensus driven high-trust societies, with similar technological concept for SNF disposal, and whose licensing processes have advanced at fairly similar pace, both also possess their own distinct characteristics. One difference concerns the role of the communities in siting and licensing of the repositories. The paper examines 1) the background for this differentiation, 2) how local final disposal organisations in the host communities Eurajoki and Östhammar took shape and evolved, and 3) how differences between the organisations illustrate the divergence between the Finnish and Swedish approaches to stakeholder engagement. While the Swedish approach can be characterised as 'involved partnership' – which shapes the operating environment for the implementer and authorities, by challenging and even modifying the policies and actions – the Finnish case could be described as a 'bystander partnership' characterised by trust in safety authorities, with community economics as the primary concern.
Canada’s Nuclear Waste Management Organization (NWMO) is leading a site selection process for an informed and willing host community with a suitable site for a deep geological repository for used nuclear fuel, as well as an associated Centre of Expertise. The process was initiated in 2010 and is expected to culminate with identification of a preferred site around 2023. It is a community-driven process designed to address a broad range of social, economic, cultural and technical factors identified through dialogue with Canadians and Indigenous peoples. The process involves a step-wise approach with clear decision points, and increasingly intensive stakeholder engagement and technical study. Consistent with the NWMO’s commitment to involving people in its work, the siting process is being implemented in an open, transparent and inclusive manner through a growing set of engagement and communications programs. These programs are frequently shaped by the very stakeholders they aim to engage, and seek to: build awareness, understanding and support among key audiences; work collaboratively to identify potential repository sites that are socially acceptable and respectful of social and cultural values; and explore potential to build supportive partnerships to implement the project while enhancing well-being and building resilience of communities. This paper provides an overview of the site selection process, with a focus on approaches used to engage and communicate across a wide range of audiences and platforms to achieve the goals described above. It explores the types of programs and activities used to engage citizens in developing Canada’s plan and the site selection process, and in implementing the project collaboratively with municipal and Indigenous communities. It also discusses how the NWMO is expanding and adapting the activities, tools and platforms it uses to increase visibility and understanding of its work among key audiences in preparation for site selection.
Challenges for safeguarding a geological repository of spent nuclear fuel pose many high-level opportunities. First, being a relative late-comer among the various types of nuclear facilities subject to safeguards, the geological repository is an ideal candidate for applying “safeguards by design” (SBD).
Second, a repository is unlike all other nuclear facilities such that containment and surveillance (C/S) arguably should constitute the primary safeguards approach, rather than material accountancy.
Several states have already invested many years and resources toward implementing final disposal of spent nuclear fuel in a geological repository. We consider the unfolding safeguards consideration of geological repositories from the perspectives of SBD, safety, security and safeguards (3S), and C/S.
In this paper, a proposed approach for efficient implementation of Safeguards-by-Design (SBD) early into the design process of repository of spent nuclear fuel is proposed. The proposed approach describes the involved parties, their roles and responsibilities, ways of coordination and collaboration as well as main areas to be considered. International best practices in this regards are also presented and discussed
It is envisaged that all spent nuclear fuel generated during the operation of the RBMK-1500 reactors at the Ignalina NPP will be stored in dry storage facilities for at least 50 years prior to its disposal into a deep geological repository. According to the Radioactive Waste Management Development Programme (approved in 2015) the construction of the repository is planned to be completed in 2066, and all SNF should be disposed of until 2073. Before the construction of the repository, various preparatory activities shall be performed: site selection, repository concept and designing, environmental impact studies, safety analysis, etc. There are two geological formations in the territory of Lithuania potentially suitable for the construction of the repository – crystalline rock and clayey formations. Dry storage casks that are currently used for RBMK-1500 SNF interim storage at Ignalina NPP site cannot be used for the disposal purposes. Therefore, SNF reloading from the storage casks into appropriate disposal canisters will be necessary. The type of the disposal canister depends on the geological formation in which the repository is constructed. According to the existing knowledge, copper canisters are considered appropriate for disposal into crystalline rock and steel canisters are suitable for clayey formations. This paper presents preliminary criticality and radiation safety evaluation of copper and steel canisters containing RBMK-1500 spent fuel. Radiation characteristics and dose rates on the surfaces of the canisters are modelled assuming SNF disposal after 50 and 100 year interim storage.
Deep Isolation brings technical innovation and creative design to the nuclear waste disposal impasse. Deep Isolation offers a solution for safe, secure, and permanent deep geological disposal of nuclear waste while reducing the time and cost of licensing, packaging and transportation. Deep Isolation uses established directional drilling technology from the oil and gas industry to drill a vertical drillhole 1 km to 2 km deep that transitions to a horizontal drillhole 1 km to 3 km in length. The target disposal media are geologic formations whose stability has endured for millions of years. Deep Isolation proposes the direct disposal of spent nuclear fuel assemblies – or other high-level waste – in specially designed canisters in the horizontal section of the drillhole. Disposal of nuclear waste in horizontal drilled holes increases the effectiveness of both the engineered and natural barriers preventing the release of radionuclides from the nuclear waste into the biosphere. The availability of this suitable geology throughout the world allows for disposal at multiple locations. The horizontal emplacement allows for appropriate spacing given the heat generation of the canisters and minimizes the potential migration of radioactive contamination. Deep Isolation is committed to listening to and learning from the public and interested parties in the pursuit of disposal solutions. As such, our approach mirrors our values of openness and transparency. At present, Deep Isolation is actively listening to and exploring public and stakeholder attitudes toward the deep horizontal drillhole disposal concept more generally. Recently, Deep Isolation completed a self-funded demonstration of its technology. A prototype canister was placed into the horizontal section of an existing drillhole. The canister was released, and the installation equipment returned to the surface. Subsequently, the recovery equipment was then sent into the drillhole, captured the canister and returned to the surface successfully demonstrating retrievability.
The immobilization of 90Sr liquid waste simulated by 88Sr(NO3)2 solution with polyethylene terephthalate matrix mixed with a polystyrene additive had been carried out. These materials were selected due to its properties that is suitable for immobilization of radioactive waste. In addition, the usage of polyethylene terephthalate and polystyrene are able to reduce plastic waste. Zeolite was used to adsorp 88Sr. Polystyrene additive was varied to 10%, 20%, and 30%. Polyethylene terephthalate and polystyrene were heated at 250°C for 20 minutes to melt. Zeolite which has adsorbed 88Sr(NO3)2 was added into the mixture and stirred to homogenous mixture. Then, the mixture was molded and cooled. Flash-fire point testing was performed with the Cleveland Open Cup whereas the penetration testing was carried out with a penetrometer. Results of the flash and fire point testing shown that there is an effect of adding polystyrene additive on the waste-polymer block. The highest thermal resistance is on the 30% polystyrene additive. The highest flash and fire point are (340.67 ± 0.58)°C and (356.33 ± 0.58)°C, respectively. In a penetration testing, there isn’t the effect of polystyrene additive because the test results 0 mm value for all variations. All of the above test results meet the specified standard (GOST/Russian’s ANSI) or better than the average value of the previous product.
The paper describes the process of forming requirements for the disposal system for radioactive waste on the example of the thermal regime of deep radioactive waste disposal facility.
The work aims to assess a novel prepared composite of Poly Acrylic Acid/Charcoal/Montmorillonite (PAACM) as a backfill material for radioactive waste disposal facilities. Characterization of the prepared composite has been done using Fourier Transform Infrared Spectroscopy (FTIR), Thermal Gravimetric Analysis (TGA), Energy Dispersive X-ray (EDX), and Scanning Electron Microscopy (SEM). Cs+ and Sr2+ ions were determined using Inductive Coupled Plasma (ICP). Some factors which may affect in the sorption process such as contact time, pH of the aqueous phase, mass of the sorbent, competing ions and temperature of the aqueous medium has been studied. The obtained data show that the sorption process is rapid and the PAACM composite has a high sorption capacity towards Cs+ and Sr2+ ions, the PAAMA has high thermal stability, and the PAACM has highly enormous swelling properties. The swelling of PAAMA is pH and temperature dependent; of the aqueous phase. K+ and Ca2+ ions are good competing ions for Cs+ and Sr2+ ions, respectively, during sorption process. Application of the principles of radioactive waste management requires the implementation of measures the activity level of radioactive waste that afford protection of human health and the environment, now and in the future. The prepared material contributes in the protection of the environment and groundwater from contamination with both radio-Cesium and radio-Strontium.
Repository induced effects (RIE) play an important role in the assessment of long-term safety of a deep geological repository for Spent Fuel and High-Level Waste (SF/HLW). Three categories of RIE and the associated indicators are assessed in this study. The generation of heat from SF/HLW canisters causes build-up of pore water overpressure in the repository near-field, whereas gas generation from the corrosion of waste and construction materials can lead to gas pressure buildup in the repository structures and surrounding rock. The third effect is related to the development of the Excavation Damaged Zone (EDZ) around the repository; the vertical extent of the EDZ is associated with a reduction of the lengths of radionuclide release paths through unaffected host rock. A generic RIE indicator approach presented in this study, providing a numerical framework for quantitative assessment of: (a) the general relevance of RIE for long-term safety of the repository system and, (b) the potential of the RIE indicators to discriminate between candidate repository sites. For the evaluation of the RIE, three-dimensional (3-D) models of the entire repository system as well as two-dimensional (2-D) representations at the component scale (e.g.; models of a single emplacement room). are developed. Scoping calculations are performed in support of the integration of repository-scale with component-scale analyses and the associated dimensionality reduction (3-D to 2-D) applied to the different models. To assess the RIE, probabilistic assessments can be integrated with the uncertainty of the entire ensemble of input parameters and the propagation to model predictions can be estimated in a reliable and computationally efficient manner. The generic workflow presented in this study is considered as a versatile tool for site-specific assessment of the RIE indicators in future site selection programmes.
The purpose of the paper is to share practical observations regarding knowledge management based on experience following the suspension of the proposed Yucca Mountain Project (YMP) in the United States. After its suspension in 2010, custody of YMP information systems was transferred to the U.S. Department of Energy (DOE) Office of Legacy Management. Meanwhile, Sandia National Laboratories was directed to maintain, on its own systems, the technical basis supporting the postclosure component of the 2008 Yucca Mountain Repository License Application. Because of increasing costs of hardware maintenance and increasing risks of obsolescence of some of the database software used at the YMP, the DOE has directed Sandia to develop a cloud-based information management system to serve as a generic template for any future nuclear waste management and disposal project. Generally, IT systems for large, long-term programs are developed incrementally, as needed, by separate organizations for their specific purposes. Opportunities to restructure and integrate these various systems are rare because of the disruption caused. This cloud effort is being developed using the experience and the information systems from the former YMP and its requirements while there is no ongoing scientific and engineering work, providing a unique information management opportunity to analyze, define, and potentially integrate an information architecture that meets all its needs and requirements efficiently. Key observations relevant to knowledge management include:
(a) Observations on opportunities and challenges of migrating information to a cloud platform.
(b) Thoughts on efficiently managing and structuring information to meet separate requirements and needs.
(c) Observations on capturing tacit knowledge from experts and using it in information management.
(d) Observations on knowledge recovered from legacy project software itself.
(e) Implications on long-term management of information and records to meet regulatory requirements and to maintain useful knowledge and lessons learned for other programs and purposes.
Science and engineering provide the necessary answer to the ultimate question in radioactive waste management and disposal: How safe is the management approach and the repository system? The credibility of that answer is founded on underlying processes and systems that demonstrate the reliability of the information used to answer this singularly important question. This technobureaucratic culture is often assumed to be effortless and is taken for granted, and assumptions like this can lead to unacceptable results.
These non-technical processes fall into two broad but related categories--regulatory compliance and information/knowledge management. In addition to specific technical regulatory requirements, in the United States (U.S.), the U.S. Nuclear Regulatory Commission (NRC) requires compliance with several abstract concepts that it views as essential to demonstrating that an organization has the requisite wherewithal to be a licensee, such as Nuclear Safety Culture, Safety Conscious Work Environment, and Quality Assurance. These concepts greatly influence all the technobureaucratic processes and systems that support the science and engineering work.
This paper presents a generic framework for an organization and the functions of the organizational elements necessary to execute a generic radioactive waste repository development effort. These organizational elements reflect a workforce’s functional composition and the practices that facilitate meeting all of the NRC’s expectations.
Successful implementation of a plan to develop a repository requires an effective organization and infrastructure designed to execute the effort in compliance with regulatory expectations. The discussions in this paper are based on the current U.S. statutory and regulatory framework. Notably, the context in which the organization’s work will be conducted differs substantially from that of the typical research, development, and demonstration (RD&D) environment. First, there are work elements that are not customarily included in RD&D work, such as regulatory compliance, a corrective action program, technical configuration controls, and requirements/commitment management. Secondly, the rigor with which organizational assurance and quality assurance functions need to be applied and practiced is greater than necessary in the typical RD&D environment.
One all-too-frequently overlooked component of a compliance-oriented endeavor is the importance of having an outcome-aware management and business organization, technical support, and information management technologies. Successfully accomplishing such an endeavor requires more than world-class science and engineering. It is equally important that the technical team be supported by an experienced and proficient non-technical infrastructure.
Operations that extend from the process of uranium ore mining to the step of reprocessing are well known as nuclear fuel cycle (NFC). NFC consists of two ultimate parts the first part is called "frond end" while the other is named "back end". The back end of the NFC involves managing the spent fuel after irradiation. IAEA executes safeguards system on sates under the non proliferation umbrella. This system ensures not only nuclear material (NM) but also the activities within facilities are subject to supervised criteria accredited internationally and supported by states acceptance. Safeguards approaches and the application of safeguards is facility specific. Implementation of safeguards includes inspection on facilities that contain spent fuel. The paper high lights the SNF signatures such as physical signature, gamma radiation, Cerenkov radiation, neutron radiation, and combined radiation. Each signature gives safeguards inspector a piece of information concerning the nuclear fuel and the process it passes through. Discussions on spent fuel safeguards and verification objectives are presented also. NMA verification objectives are to detect gross defects like missing a spent fuel assembly also to verify the identity of SNF to ensure that a spent fuel assembly is the assembly that declared by the facility operator another verification objective is to detect partial defects like verification of the integrity of SNF object. The Containment and surveillance (C/S) verification objectives are to verify continuity of knowledge over SNF assemblies and to verify no use or production of undeclared nuclear material. Design Verification objectives are to verify facility design (no new unsafeguarded SNF transfer paths). Eventually, nuclear abuse scenarios are suggested and the role of any robust accounting system in safeguarding SNF was discussed to stand up to these concealment tricks.
The main objective of the study is the theoretical estimation of long-lived actinide and fission product inventories by combining depletion calculations and non-destructive measurement techniques (measuring of gamma-ray emitting fission products). This is important on account of possible leakage from long-term disposal of spent nuclear fuel. Identification of the most radiotoxic actinides and fission products in different cooling times (especially at long-term) and their behaviour from radiological protection aspects is discussed in the study. The solubility and mobility of long-lived actinides and fission products in underground waters, rocks and soil is very important for the accurate prediction of environmental impacts in the case of leakages from long-term disposal facilities. The results of this research are expected to improve the estimation of long-lived alpha emitting actinides, which is usually measured by destructive methods.
The inhalation and ingestion radiotoxicities are calculated along with the concentration, radioactivity, thermal and gamma power, and neutron and gamma spectra in different cooling times (up to 106 years). Different parameters were varied, such as fuel enrichment, fuel temperature, burnup history, length of cycles, and local irradiation environment, in order to determine their influence on inventories. Variations of operational conditions and local irradiation environment during irradiation have a significant impact on isotopic inventories for the Halden reactor.
Investigation of inhalation and ingestion toxicities of specific actinides and fission products revealed that the larger variations are seen between 3000 – 100000 years, which is relevant time frame for the postulated leakage from the long-term disposal according to the reported results from previous studies. Preliminary results of this study have shown that the long-lived actinides, such as 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 245Cm, 236Np, 237Np, 235U and 238U and fission products, 79Se, 99Tc, 129I and 135Cs are the important radionuclides in the terms of radiotoxicity, as well as solubility and mobility.
The 6.0 % enriched Halden driver fuel irradiated up to 60 GWd/tUO2 is chosen for this study and is modelled by using SCALE-6.1 codes system. TRITON code is used for the building of Halden driver fuel specific cross section libraries and ORIGEN-ARP is employed for the depletion analysis. The 238-group ENDF/B-VII.0 neutron library is used for the calculations.
Nowadays several countries are planning to store nuclear spent fuel in long-term geological repositories. The fuel will be preserved for thousands of years inserted in copper canisters with iron inserts. In Sweden, after the encapsulation of the fuel, canisters will be welded and transported to the geological repository for the deposition in tunnels. During transport, the Continuity of Knowledge (CoK) of spent fuel must be kept. An option could be the identification of canisters. Traditional tagging methods do not fit all the requirements foreseen in this application and then alternative ideas should be developed. The method proposed by the Joint Research Centre of the European Commission, in collaboration with the University of Florence (Italy), could guarantee not only a unique label for each canister but also security against falsification attempts. The idea is to combine two fingerprints: the first (artificial signature) related to a configuration of chamfers arranged around the lid circumference and the second (natural or intrinsic signature) due the variation of the internal gap between lid and tube after the welding process. The geometry and position of chamfers are designed to maximize the ultrasonic echo without affecting too much the canister structure and minimize the influence on the mechanical properties of the canisters. The ultrasonic amplitude response (identification fingerprint) acquired by a rotating high frequency (10 MHz) transducer with a fixed height is strictly related to chamfers disposition and then it is unique for each container. However, in order to verify the originality of canisters, the internal gap between lid and tube is investigated. The ultrasonic response due to the inspection of the welding area is a “natural” fingerprint (authentication fingerprint) for canister because it is connected to the welding process and material properties. The identification and authentication of fingerprints can be combined by angular matching to increase the robustness of the method. The description of the method and the development and testing of the ultrasonic reading system are reported in the paper.
For the disposal, intermediate storage and transport of spent nuclear fuel a number of properties of each fuel assembly must be determined, both for operational and safeguards needs. Important examples of these parameters are decay power, multiplicity, burn-up (BU), initial enrichment (IE), cooling time (CT), completeness of fuel assemblies, weight, amount of fissile material and nuclide inventory. This is done through a combination of known fuel history, measurements and codes.
In addition, the status of the fuel assemblies is necessary to characterize. Failed or damaged fuels must be identified prior to final disposal in order to treat them appropriately, as are other mechanical and chemical issues that may affect the handling in the system.
The uncertainties of these determinations are crucial in the use of the parameters, and are judged to be fairly large at present. Particularly the uncertainly of the decay power has a direct relationship to the cost of any repository due to temperature requirements in the systems. These cost savings are potentially very high, in the order of billions of Euros. A thorough understanding of these issues also opens ways to optimize the facilities, for example economically and environmentally.
Due to the large amount of fuel assemblies to be measured, high through-put and robustness of the methods and instruments are paramount, as is the capacity to make fast decisions made on the measurement results and codes.
The status and future needs of development of instruments, basic fuel data and cross sections, and codes will is discussed in the paper, and how this is done in various collaborations world-wide. Potential problems, such as errors in fuel data, uncertainties in basic nuclear data, uncertainty propagation, conflicting methods and results etc., is illustrated and discussed.
An international effort to blindly test the capacity to calculate decay power on fuel history, led by SKB and in collaboration with NEA/OECD – with more than 25 participating organizations and groups, using most of the internationally available codes, is described.
Finland was the first country to license a spent nuclear fuel encapsulation plant and a disposal facility. Posiva Ltd (Posiva) submitted the construction license application (CLA) in the end of 2012. Radiation and Nuclear Safety Authority (STUK) made the review of the CLA and in February 2015 the statement and safety evaluation report were submitted to the Ministry of Economic Affairs and Employment (MEAE). The construction license for Posiva was granted by the Government in November 2015.
STUK started planning the regulatory control of the construction activities after finishing the CLA review. The regulatory control of the encapsulation plant construction follows mainly the principles of a nuclear facility control taking into account the graded approach. The control of the construction of the underground disposal facility requires a novel approach since excavating safety classified rooms in the bedrock keeping in mind the long term safety aspects differs from the other types of nuclear facility construction.
Posiva constructed the underground rock characterization facility, Onkalo, at the facility site during 2004 - 2016. STUK controlled the construction of Onkalo similarly as it would have been a nuclear facility because from the very beginning Onkalo was planned to be a part of the future disposal facility. The experience gained from the control of the Onkalo construction was applied in the development of the regulatory control concept of the disposal facility construction.
The article will give an overview on the regulatory control concept of the spent nuclear fuel encapsulation plant and the underground disposal facility during the construction phase. The construction phase covers the detailed design of systems, structures and components, as well as the feasibility demonstrations, the construction activities, monitoring the construction effects to the underground disposal facility and the commissioning.
The review phase of the operating license application is expected to begin in the early 2020s. The planning for the review of the operating license application is on going at STUK. This aspect is also discussed in the article.
In the Russian Federation, more than 6,000 tons of vitrified HLW from the reprocessing of SNF have been accumulated, which are planned to be disposed of in a deep geological repository. Spent nuclear fuel composition is determined by various types of processed SNF (Russian nuclear power reactors VVER-440 and BN-600, research reactors, etc.), technological features of processing and cooling period. Accounting these factors with accuracy required for the safety assessment of the disposal both during the operation period and after the closure of the storage requires the development of special approaches, methods, etc., including a wide set of mutually agreed measurements, radiation characteristics and computational studies to predict the activity of RW and to reduce uncertainties for packages with RW, physics and chemical characteristics of vitrified RW.
To confirm the simulation results, it becomes necessary to verify and validate the code for calculating the nuclide composition of SNF based on a comparison of the calculation results and reference experiments. The international experts work on the selection and compilation of integral experiments to determine the nuclear-physical characteristics of SNF and have published their work to determine the radiation characteristics of SNF and the accumulation of fission products and actinides in it, including for the SNF from VVER-440 reactors.
The report focuses on planning and justifying work on obtaining HLW compositions (based on the analysis of experiments), information on which is the foundation for planning an activity on vitrified HLW.
The question of disposing of radioactive waste after it has been generated is an ongoing issue for the nuclear industry. Currently one of the preferred solutions is to encase the waste in containment structures and bury it deep underground until the radioactivity has decayed to safe levels. In order to prevent future human intrusion, the repositories containing the waste much be clearly marked in a way that understandable for future society.
The paper covers the previous research efforts to develop a suitable warning system for informing future generations of the hazard posed by radioactive waste interred in a deep geological repository (DGR) or geological disposal facility (GDF) and discusses the merits a variety of approaches as well as the ethical considerations of building such a system.
Various roadmap exercises have been undertaken during the past 2 decades relating to the development and deployment of 'advanced' nuclear energy systems, particularly so-called "Generation-IV", to improve the sustainability of nuclear energy. While nuclear energy is already among the most sustainable energy conversion technologies, the spent fuel (SF) management particularly remains a major socio-political challenge to further the use of nuclear energy in sustainable energy mixes. The availability of natural resources being, for the time being, less a driver towards such 'advanced' sustainable nuclear energy systems.
Though, such advanced nuclear energy systems require competitive technical-economic performance while addressing the socio-polical challenge for improved SF-management and, in today's energy market and socio-political environment, there's (very) limited willingness by private sector and even by most governments to embark on the effective deployment of such advanced nuclear energy systems.
Three categories of 'advanced' nuclear energy systems are considered and analysis in this paper:
a. nuclear energy systems using more advanced synergies between existing or near-term deployable nuclear technologies and where the development effort as such is limited but where especially international win-win situations need ot be recognised and deployed by mostly the private sector though considerable for deployment during the next 2 decades;
b. "Generation-IV'-systems, mainly HTGR and FRs, under development internationally and providing essentially multi-recycling options of separated materials in addressing SF-management challenges. These systems potentially seeing market deployment from the 2040s on in synergy with the LWR/PHWR-park worldwide;
c. "Generation-X"-systems focusing on very advanced nuclear energy systems and particularly seeking transmutation avenues aimed at furthering the reduction of transuranics amounts to be disposed of. These systems, from a technical-economic perspective, only deployable well after mid-century.
This paper presents an analysis of the technical-economic effectiveness of these three categories of advanced nuclear systems in transitioning towards more sustainable nuclear energy systems. A multi-regional analysis, reflecting real data on nuclear energy systems evolution in these regions, mapping the time-lines to address the socio-political challenges for the (regional) transition towards more sustainable nuclear versus the (regional) technical-economic performance challenges is presented accordingly.
Partitioning and Transmutation technology is expected to be effective to mitigate the burden of the high-level waste (HLW) disposal by reducing the radiological toxicity and heat generation. Based on the Strategic Energy Plan of Japan, research and development (R&D) on P&T are being accelerated in Japan. The Japan Atomic Energy Agency (JAEA) has been continuously implementing R&D on P&T technology. The R&D on P&T in JAEA are basing on two kinds of concepts: one is the homogeneous recycling of minor actinide (MA) in fast reactors and the other is the dedicated MA transmutation, so-called “double-strata” strategy, using an accelerator-driven system (ADS). The ADS proposed by JAEA is a lead-bismuth eutectic (LBE) cooled fast subcritical reactor with thermal output of 800 MW. Various R&D activities not only for ADS but also for advanced fuel cycle are progressing in JAEA. For the partitioning process of minor actinide from spent fuel, a new separation process, SELECT process, using new innovative extractants to improve the partitioning process from the viewpoints of the economy and the reduction of secondary wastes was developed. For the MA-bearing fuel for ADS and its fuel cycle are also being developed. Uranium-free nitride fuel was chosen as the first candidate for ADS. A pyrochemical process has been proposed as the first candidate for reprocessing of the spent nitride fuel. In this paper, recent R&D activities based on these policies are briefly shown.
The current open nuclear fuel cycle uses only a few percent of the energy contained in uranium. This efficiency can be greatly improved through the recycling of spent fuel (as done today in France for instance), including, in the longer term, multi-recycling strategies to be deployed in fast reactors. In this context, and in the continuity of the FP7 EURATOM SACSESS project, GENIORS addresses research and innovation in fuel cycle chemistry and physics for the optimization of fuel design in line with the strategic research and innovation agenda and deployment strategy of SNETP, notably of its ESNII component. GENIORS focuses on reprocessing and fuel manufacture of MOX fuel potentially containing minor actinides, which would be reference fuel for the ASTRID and ALFREDO demonstrators. More specifically, GENIORS carries out research and innovation for developing compatible techniques for dissolution, reprocessing and manufacturing of innovative oxide fuels, potentially containing minor actinides, in a “fuel to fuel” approach taking into account safety issues under normal and mal-operation. The different promising options developed in SACSESS are currently further developed to address the specific challenges of GEN IV. For delivering a full picture of a MOX fuel cycle, GENIORS works in close collaboration with the INSPYRE project on oxide fuels performance. By implementing a three step approach (reinforcement of the scientific knowledge => process development and testing => system studies, safety and integration), GENIORS will lead to the provision of more science-based strategies for nuclear fuel management in the EU. It will allow nuclear energy to contribute significantly to EU energy independence. This paper presents the strategy and current results of GENIORS.
Closed nuclear fuel cycle (CNFC) with inherent safety fast reactors (FR) is a new integrated product in the branch of atomic energy. A pilot demonstration power complex with reactor unit (BREST-OD-300) with lead coolant is under constriction at the Siberian Chemical Combine in frame of the “PRORYV” project. This pilot demonstration power complex includes not only fast reactor but also facility for fuel fabrication/refabrication and facility for mixed uranium-plutonium nitride (MNIT) spent nuclear fuel (SNF) reprocessing and radwaste management. Integrated system of models and codes for the coordinated simulation of different processes and phenomena for CNFC technologies are also under development.
India has adopted a ‘closed fuel cycle’ considering spent fuel a material of resource. This has enabled not only optimally utilising the scarce resource of Uranium but also helped in efficient management of radioactive-waste and opening the possibilities for tapping the energy of various useful radio-isotopes present in waste for societal benefits which otherwise are not available in nature. Reprocessing of spent fuel enables in recovering of fuel and recycling them to future reactor for utilising as fuel. Such recovery and recycling of fuel material in reactor to generate power not only helps in ensuring the energy security of the country but also helps in reducing the rad-waste volume meant for geological disposal to a great extent. Spent fuel reprocessing results in recovery of more than 95% of material and hence generates a very small amount of high level liquid waste (HLLW) which is significantly lower than the direct disposal of spent fuel in case of ‘open fuel cycle’.The HLLW, characterised by high concentration of radioactivity in combination with presence of long lived minor actinides, poses the challenge for its safe management. The HLLW is vitrified in suitable glass matrix and interim stored for removal of decay heat. The advantage of vitrification of HLLW into vitreous matrix is to immobilise the radioactivity in chemically durable form ascertaining containment and isolation of radioactivity from the human environment for extended period of time.
HLLW contains many valuable radio-nuclides such as Cs-137, Sr-90, Ru-106 etc which have various societal applications in the field of industry and healthcare. India has put a step forward in implementation of advance fuel cycle by recovering the valuable radionuclides from HLW and deploying them for various societal applications. Separation science has played a key role in selective recovery of these radio-nuclides in pure form from HLLW. Recovery of Cs from HLLW using solvent extraction based system enabled use of Cs in non-dispersible glass form for irradiation purpose. Recovery of Sr-90 from HLLW was also demonstrated to milk out the radio-pharmaceutical grade Y-90 for radiopharmaceutical applications. Recently, Ru-106 has been recovered from HLLW to produce Ru plaque for eye cancer treatment.
Reprocessing of spent fuel for recovery of heavy metals followed by extraction of useful radioisotopes reduces the waste volume immensely prior to isolation and their eventual disposal. The paper outline, the practices being adopted in India for management of high level radioactive waste. A brief description covering the important aspects of waste management like the sources of HLW, composition details along with management strategy is given.
The current direction of energy politics adopted by the Japanese government is the utilization of nuclear energy, therefore, the argument of radioactive waste management is becoming increasingly important. To promise a safe, and less environmental load disposal repository for the high-level radioactive waste (HLW), it is necessary to look ahead of future fuel cycle system which intends to use plutonium by the introduction of mixed oxide (MOX) fuel together with the extended cooling period of UO2 spent nuclear fuel (SF), which are not much considered until now. If such fuels are reprocessed and vitrified in near future, it will have much impact on the heat generation of the vitrified waste arisen from 4-year-cooled UO2 spent fuel from discharge and will lead to an increase in the footprint of the geological repository.
To reduce the volume of HLW and the footprint of the geological repository, partitioning technology is considered as an option to help to solve the issue. By undertaking partitioning, separation of the heat generating and radiologically harmful nuclides such as minor actinide (MA: Np, Am, Cm) from HLW is expected and the related development of technology is promoting for the advanced nuclear fuel recycling. Then, it leads to higher waste loading ratio of vitrified waste and consequently the less space needed for final disposal due to the less amount of heat generation from vitrified wastes. In this study, the evaluation of effects of MA separation on spent MOX fuels and spent UO2 fuels with prolonged cooling period were performed since one of the major factors hindering the higher loading is 241Am and its decay heat affects adversely to the geological repository. Therefore, heat transfer calculation was carried out to evaluate the temperature of buffer material in a geological repository. For the UO2 SF which cooling period is longer than 50 years, the reduction of the maximum temperature of the surface of the buffer material due to the MA separation was large, and it was lower than the upper limit of buffer material temperature of the surface, 100℃, by 70% separation of MA. This indicates that when the cooling period of SF is prolonged, MA separation impacts more on the reduction of the surface temperature of buffer material. However, the introduction of MA separation was not sufficiently effective in terms of the thermal property of a repository for spent MOX fuels because the amount of 241Am was increased by beta decay of 241Pu. Through this study, the relation among fuel type, cooling period of SF, waste loading, MA separation, and the disposal repository was revealed.
The Th fuel cycle is attracting interest again globally because of its advantages over the current Pu fuel cycle, such as breeding fissile 233U from fertile 232Th without using a fast reactor, lower minor actinide production and higher Pu burning. However, there are some concerns, such as the small critical mass of the bred 233U. Using 234U, which is not considered an important isotope, may overcome some problems with the Th fuel cycle. In this study, the effect and roles of 234U in the Th fuel cycle were surveyed from the perspectives of proliferation resistance (PR), fuel burn-up, and breeding in single and multiple cycles. Increasing the 234U isotope ratio increases bare critical mass, which in turn increases PR by increasing the heat generation and radiation dose rate from 232U and their daughter nuclei. The effects of the moderator-to-fuel ratio, neutron energy spectrum, and neutron flux (linear power density) on criticality were estimated. 234U was fissile in the faster neutron energy spectrum, which can increase the fuel burn-up under some conditions. A higher fuel burn-up is preferable to increase the 234U isotopic ratio. For multiple cycles, the breeding ability of 234U was higher with a softer neutron energy spectrum (33.3% at the end of the fifth cycle), but the mass balance was worse. When 234U was used with a harder neutron energy spectrum, the 234U isotopic ratio was as high as 23.6%, but the mass balance was better. The role of 234U in Th has not been thoroughly investigated until now, but this study has revealed the importance of 234U, which may lead to the development of a new Th fuel cycle.
Minimizing the volume of nuclear energy waste becomes more and more vital issue year by year. In many countries, the solution of this problem is a corner stone for taking decision on new nuclear build and for dilemma either to extend the national nuclear energy program or to phase out. Rosatom pays special attention to development of technologies aimed at activity and volume reduction for radioactive waste (including HLW) which is subject to final disposal, especially in terms of fractioning HLW resulting after spent nuclear fuel processing.
As Indonesia is planning to build an experimental power reactor of pebble bed HTGR type, it is important to determine the content of important fission product nuclides of its spent fuel. Identification the amount and type of the FPs is the first step toward the implementation of safeguards policy, management of the spent fuel and addressing the source term strength in case of an accident. The calculation was done using Monte Carlo method MCNP6 Code. It is intended to calculate the amount of nuclides that are important to safeguards, such as the remaining U-235 and the produced Pu-239; the amount of long-live minor actinides that are subject to spent fuel management; the amount of fission products that are important in addressing radioactive release in case of an accident.
Rare earth (RE) oxides are one of the waste streams generated during pyroprocessing of spent nuclear fuel. To immobilize RE oxides in glass, we explored an international simple glass (ISG), alkali borosilicate glass (ABG) and aluminoborosilicate glass (AG). The loading of RE oxides varied within 15 and 56.5 mass%. To characterize glass durability, we performed the 7-day Product consistency test (PCT), which showed good values for ISG and AG glasses. Crystallinity was investigated using optical microscope and x-ray diffraction (XRD). Various types of crystals were identified: ISG formed RE-borosilicate crystals with a single lanthanide and oxyapatite with a mixture of several RE oxides; ABG formed cerianite; and AG formed cerianite, RE-borosilicate, and Al-containing crystals. We have determined liquidus temperatures for crystals formed as functions of glass composition.
Current status of the nuclear fuel cycle in Russia is characterized by significant increase in NPP units under construction based on reactor facilities with VVER-1000/1200/TOI. At the same time rapid expansion of the reactor fleet faces the back-end problems that have to be dealt with. These problems concern reprocessing of SNF and management of products generated as a result of the reprocessing.
The closed nuclear fuel cycle is a basis of package of services offered by Rosatom State Corporation at foreign markets.
The thermal properties and amount of vitrified waste are major factors that determine the eventual disposal area of high-level radioactive waste deep underground. The effect of high burn-up operation of a light-water reactor with UO2 fuel on the amount and thermal properties of vitrified waste under various nuclear fuel cycle conditions was discussed. In addition, the effect of Cs and Sr separation and high-content vitrified waste on reducing the waste-occupied area, which may affect the geological disposal area, under high burn-up conditions was quantitatively evaluated by using the Comprehensive Analysis of Effects on Reduction of disposal Area (CAERA) index. The fuel burn-up had a limited effect on the amount of vitrified waste. Furthermore, the contribution to the heat generation rate of vitrified waste for high burn-up conditions of 137Cs, 90Sr, and their daughter nuclides, which have relatively short half-lives, increased and contribution of 241Am, which has a longer half-life, decreased. Therefore, high burn-up conditions reduced the waste-occupied area via Cs and Sr separation, and the maximum effect was a reduction of 74% of the waste-occupied area with a fuel burn-up of 70 GWd/tHM, 4-year spent fuel (SF) cooling period, 90% Cs and Sr separation, and 30 wt % vitrified waste loading. The results suggested that fuel burn-up, SF cooling period, partitioning technology, and vitrified waste loading are important for the geological disposal area, and it is necessary to consider the combination of these conditions for reducing the geological disposal area.
This contribution is supported by a poster and also an oral presentation. The full paper is included in the Conference with ID#34.
The paper presents the main Russian concepts and projects in the field of spent nuclear fuel and high level liquid waste reprocessing. A new concept of partition was proposed to minimize waste requiring deep geological disposal. The key problems and tasks of the concept are discussed.
Clean energy is most desirable in the world without any problem; nuclear energy can fill the thirst of energy with lower problems. Waste disposal and radiotoxicity is the most challenge issue in nuclear power generation. It is suggested to transmute spent nuclear fuel and especially minor actinide due to it is high contribution of radiotoxicity. We transmute the Np-237 in the LWR AP1000 reactor. We simulate the reactor and the minor actinide materials using mcnp5 code. We got promised and good results to recommend realizing this issue.
The mathematical model of uranium extract purification process from technetium was made in TPU in the CODE-TP media. In order to maximize the adequacy of the CODE-TP calculations, experimental verification of the simulating algorithms for the technological process of the uranium extract purification from technetium using lab-scale liquid chromatography unit (LCU) at JSC VNIINM was carried out. To test the process simulation algorithms, a series of experiments was carried out with the aim of producing data for comparison with the mathematical model
The accident at the Fukushima-Daiichi plant in Japan in 2011 highlighted vulnerabilities in the current zirconium (Zr) alloy clad uranium dioxide (UO2) fuel to an extended loss of cooling. Improving the resilience of the fuel and cladding is considered a high priority for the nuclear industry and has resulted in significant research into the development of so-called Accident Tolerant Fuels (ATF). ATF are widely expected to be deployed in the near future in existing and future Light Water Reactors (LWRs). Post discharge management and dispositioning of spent ATF is a topic that must be addressed in order to demonstrate responsible management of the fuel cycle and yet has received little attention to date. In this review the spent fuel management considerations of several leading ATF fuel and cladding concepts are assessed against current LWR fuels. The concepts include coated Zr alloys, advanced iron alloys and silicon carbide composite claddings and advanced UO2 and high uranium density fuels. Technical challenges regarding each different material are highlighted; particularly focusing on reactivity and durability in water. Recommendations are made where variations of current storage procedures are likely to be required.
Partitioning and transmutation technology will be a promising technology to reduce the burden of high-level waste disposal problem. The Japan Atomic Energy Agency (JAEA) has investigated accelerator-driven system (ADS) to transmute minor actinides (MAs) in the transmutation cycle and obtained various results. On the other hand, the neutronics design of the ADS had used the ideal fuel composition which did not include impurities such as rare earth elements or uranium.
This study aims to investigate a new ADS fuel composition based on a separation process called SELECT (Solvent Extraction from Liquid-waste using Extractants of CHON-type for Transmutation) process developed for extracting U, Pu, and MAs from dissolution solution of spent nuclear fuels and/or high-level liquid waste. By performing the neutronics calculation of the ADS with the new fuel composition, it is confirmed that the new fuel composition based on the SELECT process is acceptable. These results also indicate that the decontamination factors for RE in the reprocessing process are also adequate.
The main development of high level liquid waste (HLLW) partitioning in China was briefly reviewed. Chinese high level liquid waste has been stored for several decades. How to manage the historic HLLW is a serious problem. A total partitioning process has been developed at Tsinghua University. The process consists of the following three extraction cycles: actinides removal by TRPO extraction, Sr-90 removal by dicyclohexyl-18-crown-6-ether extraction and Cs-137 removal by calix[4]arene extraction. Based on the partitioning process, the hot test facility including 72-stage miniature centrifugal contactors was set up in the hot cell. About 300Ci HLLW was continuously partitioned within 160 hours through this facility. After 120-hour operation, 30%TRPO-kerosene was recycled without any treatment. The average values of decontamination factor were determined to be more than 3103 for activity and more than 104 for Sr-90/Cs-137, respectively. These results demonstrate that Chinese historic HLLW can be transferred into non- and intermediate/low level waste by the total partitioning process.
On the other hand, developing nuclear energy has been chosen as one of important direction for energy resources in China. The development of nuclear energy cannot be separated from the support of the nuclear fuel cycle. How to effectively manage the nuclear fuel cycle, especially high level waste from the commercial reprocessing plant, to support the sustainable development of nuclear energy is an intractable problem in China. Partitioning of HLLW provides an option to reduce the high level waste which needs be disposed in the geological repository. The research work on partitioning of commercial HLLW is under process in China.
This paper mainly considers the MOlten Salt Actinide Recycler & Transmuter (MOSART) system without U-Th support fueled with different compositions of transuranic elements from VVER 1000/1200 used nuclear fuel (UNF). Last developments concerned single fluid MOSART design addresses advanced large power unit with main design objectives to close nuclear fuel cycle for all actinides, including Np, Pu, Am and Cm. The optimum spectrum for Li,Be/F MOSART is fast spectrum of homogeneous core without graphite moderator. The effective flux of such system is near 1x1015 n cm-2 s-1. Single fluid 2.4 GWt MOSART unit can utilize more than 250 kg of minor actinides per year from VVER 1000/1200 UNF. The main attractive features of MOSART system deals with the use of (1) simple configuration of the homogeneous core (no solid moderator or construction materials under high flux irradiation); (2) proliferation resistant multiple recycling of actinides (separation coefficients between transuranic (TRU) and lanthanide groups are very high, but within the TRU group are very low); (3) the proven container materials (high nickel alloys) and system components (pump, heat exchanger etc.) operating in the fuel circuit at temperatures below 1023K, (4) core inherent safety due to large negative temperature reactivity coefficient (-3.7 pcm/K), (5) long periods for soluble fission products removal (1-3 yrs). The fuel salt clean up flowsheet for the Li,Be/F MOSART system is based on reductive extraction in to liquid bismuth. The paper has the main objective of presenting the transmutation advantages and fuel cycle flexibility of the large power Li,Be/F MOSART system while accounting technical constrains and experimental data received in this study. The main design choices and characteristics for MOSART concept are explained, including fuel maintenance and engineering safety features. The need for the experimental small power test MOSART unit to demonstrate the control of the reactor and fuel salt management with different minor actinides loadings for start up, transition to equilibrium, drain-out, shut down etc. with its volatile and fission products, is discussed
In the geological disposal of high-level radioactive waste, the amount and properties of waste buried deep underground depends on various process conditions, starting with nuclear fuel. To reduce the environmental load of geological disposal, we evaluated the effects on vitrified waste heat generation of fuel burn-up, spent fuel cooling period, separation of heat-generating nuclides, and waste loading of vitrified waste, which are closely related to the design of the repository. The timing of the appearance of the maximum temperature of the buffer material in the repository depends on the heat-generating nuclides, including short-lived nuclides Cs and Sr and their daughter nuclides and long-lived minor actinides contained in vitrified waste. The generation and accumulation of these nuclides is related to the fuel burn-up and the cooling period of the spent fuel. We examined the effect of the minor actinide separation ratio, which is related to the spent fuel cooling period, on the waste-occupied area in repository to reduce the amount of vitrified waste with high loading, assuming that molybdenum and platinum group metals were separated. Based on these results, the idea of cross-sectoral and integrated research on spent fuel, nuclide separation, vitrified waste, and geological disposal was examined to present technical options that contribute to load reduction in geological disposal.
Since 2013 the U.S. Department of Energy’s Office of Nuclear Energy began developing the Execution Strategy Analysis (ESA) tool that is both a subject matter expert elicitation process and a dynamic simulation modelling capability for use in the analysis of alternative implementation strategies and plans associated with an integrated nuclear waste management program. Early ESA models were used to evaluate potential alternatives for deploying consolidated interim storage for commercial Spent Nuclear Fuel (SNF). There have been several iterations of the ESA tool since 2013. In 2017 the ESA model was further enhanced by developing a stand-alone ESA Origin Sites Readiness Model. This model represents all the activities and milestones necessary to establish at-reactor and near-reactor site transportation infrastructure. By complementing the main ESA model and other Integrated Waste Management logistics tools, this new stand-alone model provides a structured, systematic methodology for evaluating potential SNF transportation campaigns associated with comprehensive disposition strategy alternatives.
Orano TN the logistics division of Orano has developed a Global Acceptance model. The presentation will elaborate on the methodology and give some example of how the GA model has been used for new build equipment and international transports
Communication tools and speed have drastically changed compared to 20 years ago.
Many stakeholders from national to local levels can raise questions and reluctances that will slow down or potentially shutdown the new projects.
Consequently, when implementing a new nuclear site (new build, disposal and/or interim storage site for waste and spent fuel…) and/or transporting nuclear materials, a strong global acceptance program must be developed with all the stakeholders including public in order to succeed. A high standard of professionalism must be demonstrated to all the stakeholders understanding that it is not all about the safety of the technical solutions but the way we engage and communicate.
Additionally, such program must be implemented from the initiation of the project and shall continue through the operations of the project.
Orano TN has been successful for more than 50 years in shipping various types of nuclear materials including very sensitive materials. Such success would not have been possible without a proven global acceptance model. Using a consistent and well-established methodology, such model can apply to a new site project as well to a transport project. Lessons learned over the years have allowed us to perfect our methodology.
The paper will describe the main steps of our global acceptance methodology and provide two industrial examples to demonstrate the benefit of the program.
The first example is linked to transport activities and more precisely maritime transportation of MOX Fuels and Vitrified residues from Europe to Japan.
The second example is, linked to the development of a centralized interim facility in the USA.
Addressing various items for a long term Global Acceptance program, the presentation will elaborate on these examples to demonstrate the necessity of deploying an efficient GA program to secure any investment in the nuclear field and associated transports.
In the United States, an approach to manage the aging of spent fuel dry storage systems was created by contributions from the regulatory body, storage facility owners, cask vendors, and the engineering community. The U.S. regulations for storing spent fuel beyond the first approved storage term require aging management activities to ensure that materials degradation will not adversely affect the safe storage of the spent fuel. Several guidance documents provide recommendations for complying with this regulation. The U.S. Nuclear Regulatory Commission (NRC) and the Nuclear Energy Institute (NEI) developed NUREG-1927 and NEI 14-03, respectively, to describe methods to identify the components that support a safety function, to evaluate the aging mechanisms could affect safety, and to establish aging management activities. The NEI guidance also introduces a new system to share operating experience through an Institute of Nuclear Power Operations database. The NRC also developed NUREG-2214 to identify the credible materials aging mechanisms for several cask designs used in the United States. NUREG-2214 also provides example aging management programs that may be used to effectively manage aging. Those programs rely, in part, on consensus codes and standards for monitoring and inspection guidelines, such as American Concrete Institute codes and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. Finally, to provide oversight of aging management activities, the NRC is developing internal procedures to evaluate, through inspection, the storage facilities’ performance of their aging management programs. Lessons learned from NRC Temporary Instruction TI 2690/011 will inform the development of a new NRC inspection procedure.
Fukushima drew attention to the problem of “what to do with the spent nuclear fuels that are generated, stored, accumulated, or “stranded” at nuclear power plants?” This problem is not confined only to Fukushima, Japan, but is faced by all nuclear power programs. The debate on spent fuel management has centered on what to do with plutonium, a by-product of nuclear fission. Countries concerned with the possibility of plutonium misuse prefer a once-through approach including the direct disposal of spent fuel, while others considering plutonium as an energy resource advocate reprocessing and recycling.
Regardless of how spent fuel is managed, final disposal of spent fuel and/or radioactive wastes is necessary. The challenges with disposal are complex involving technological, political, and societal aspects of finding appropriate disposal sites. An example of this difficulty was the US repository program at Yucca Mountain, which was abandoned in 2009 after 20+ years of developmental effort. Despite the challenges, Finland and Sweden succeeded in locating their disposal sites and continue to develop repositories to dispose of their respective spent fuel.
In future global nuclear energy expansion, many emerging nuclear countries would like to have an assurance of fresh fuel supply, free of disruptions to fuel their reactors. Such assurance is emboldened by the IAEA fuel banks in Ust-Kamenogorsk, Kazakhstan and Angarsk, Russia. For the back-end it is equally important to provide assurance that the spent fuel could be managed properly. Such assurance would include spent fuel take-back/away, centralized/regional storage, advanced processing technologies, and multinational repository. These back-end provisions by cooperative consortia and endorsed by IAEA are essential in forming a resilient fuel cycle, which would decouple the power generation from long-term spent fuel management, enhance nuclear safety, reduce security and proliferation risks, as well as provide flexibility and retain options for future strategic changes.
A resilient back-end fuel cycle requires innovative technologies, which include novel material development for transportation, aging, and disposal (TAD) containers; proliferation-resistant reprocessing; and alternate disposal concept such as deep-bore holes. It also needs cooperative institutional frameworks to facilitate spent fuel take-back/away; centralized/regional storage; and multinational repository. This paper examines the technological and institutional requirements and shows how the resilient fuel cycle could provide flexibility and preserve options for spent fuel management, as well as support for future global nuclear energy expansion.
Nuclear infrastructure development is vital for the effective implementation of the nuclear energy programme (including the construction and maintenance of nuclear power plants) because it helps to minimize the potential risks that may arise during the planning, construction and operation of nuclear facilities. The report addresses key issues in the development of national policies, strategies, legislation, regulation, technology, management and personnel. It presents the main factors influencing the choice of strategy in the field of nuclear fuel cycle, predicts the key directions of its long-term development and provides recommendations for the newcomer-country. The focus of the report is devoted to the back-end of the nuclear fuel cycle, the scenarios of its development and criteria for the final selection of either option. Spent nuclear fuel treatment and the possibility of returning fissile materials back to the nuclear fuel cycle determine today the whole picture of the nuclear fuel cycle.
The various options for the management of spent fuel (SF) from nuclear power reactors is a topic that has been debated from multiple dimensions, being it the socio-political concerns with regard to geological disposal, the technical-economic competitiveness of options such as reprocessing and recycling, as well as from the growing discussion on sustainability and international policy.
Many of the discussions relating to SF-management have historically been rather binomial between, on the one hand, the socio-political concerns on the direct disposal of SF and the proliferation concerns regarding reprocessing, and, on the other hand, the uncertain costs of such disposal facilities versus the economics of reprocessing and recycling schemes. Especially since the 1990s, various intergovernmental and national organisations initiated studies on very advanced SF-management schemes such as separation and transmutation also impacting the progress towards a proper solution-oriented and responsible and above-all timely SF-management.
After some decades of - generally - indecisiveness on SF-management, and with nuclear energy increasingly in the spotlight in the context of sustainable energy mixes, a more solution-oriented and responsible SF-management becomes necessary, if not urgent.
Especially as the uncertain costs and timing for such SF-management become increasingly translated into financial risks for the SF-owners, i.e. utilities. Many discussions on SF-management options were in the past coloured by strategic reflections on Unat availability and pricing, sustainable nuclear fuel cycle options (including Generation-IV systems [1]) and political considerations regarding non-proliferation. Today, there is a growing financial risk presented to utilities which becomes a more compelling trigger towards a decision on various SF-management options.
This paper addresses the changing market context for nuclear energy and particularly how SF-management options are increasingly assessed in such uncertain futures. Cost/risk optimising SF-management schemes are crucially important for utilities not to have SF as such remaining a hurdle for the future of nuclear energy’s use.
The obligation of used nuclear fuel (UNF) management lies with the NPP operator, and further – with the country which origins it. International Agreements like IAEA Joint Convention and the EU Directive 2011/70/Euratom all state that final wastes should be disposed of in the country where it is generated. Therefore, the ultimate responsibility for the management of radioactive waste (RW) lies with each member country. In many cases for the sole country and especially the sole NPP operator it is not easy to keep that responsibility: for example, for countries with small nuclear programms it will be difficult to have enough resources available for managing spent fuel and radioactive wastes. In certain circumstances, safe and efficient management of UNF and RW might be fostered through agreement among countries to use facilities in one of them for the benefit of the others.
Another potential approach to be considered is to put responsibility for UNF disposal on the fuel supplier in the model of fuel leasing. The ability to transfer responsibility for all the fuel issues including supply of the fresh one and treatment of the irradiated one is a long-standing desire of the most NPP operators , since the main task for the operator is safe and economically effective electricity production. And other tasks seem to be forced or indirect. But this desire can be hardly realized for some reasons. One of them is RW return issue: desired leasing scheme does not include the return of any RW after UNF reprocessing to the country where this UNF is generated, which, as a rule, is unacceptable for political reasons.
History knows cases when, despite to the above restrictions, the nuclear fuel leasing was implemented (for example take-back service between the USSR and Eastern Europe countries, when the used fuel of power reactors came back to USSR from GDR, Hungary, Bulgaria etc). However, these cases are the exception rather than the rule.
At the same time, the leasing concept has a number of undeniable advantages, the most important of which is the focus on the service nature in the cooperation between the supplier and the customer. The report describes how the service principle may be implemented to the fuel supply and UNF management, taking into account the current capabilities and limitations.
It is important to note that the paper does not propose any new technical solutions and inventions, but concentrates on the model that could integrate currently existing and developing solutions to be attractive to the UNF owner.
The paper analyzes different strategies for funding the disposal of Spent Nuclear Fuel (SNF). The approach is based on the idea that back-end liabilities should be funded entirely from the cash flow generated during the operation of the nuclear power plant (NPP); future generations should not be burdened with paying the costs of managing spent fuel that was used to benefit earlier generations. The framework underlying this paper is a simple one, assuming a ‘fixed-price’ world with no inflation or cost escalation over the period of NPP operation and Spent Nuclear Fuel (SNF) disposal implementation. Two key concepts used in the model are (i) the target value for the fund at the end of the NPP operation and (ii) contribution schedule - a profile of deposits into fund over the NPP operational stage. Failing to estimate these parameters correctly would lead to the mismatch of fund against liability. An important way to reduce this risk is constant recalibration, i.e. regular revisiting of expected target value of the fund and the amount accumulated over the previous periods. One of the possible strategies respecting the inter-generational equity is a contribution schedule based on constant and ongoing contributions during a station’s operating life (such a contribution schedule may be derived using a Sinking Fund Factor). The paper provides illustrative examples of one-off and ongoing contributions as well as reviewing the evolution of the fund over the duration of the NPP operation and waste management programme implementation phases. Finally, the conceptual overview of the funding strategy in a fixed-price world is introduced.
Recently, many countries start to promulgate their nuclear program all over the world. The negative issue facing those newcomers is public apprehensions about the management of nuclear and radiological activities, that any failure in their management may cause harmful radiological consequences affect directly to human health and environment. The misunderstanding of the public is considered a big challenge especially in the first stage of establishing NPP. Therefore it is very important to increase public awareness towards the management of nuclear and radioactive materials especially the more dangerous materials nuclear spent fuel(NSF) and high-level radioactive wastes(HLRWs) during storage, operation, disposal, and transport. Public should persuade that the national nuclear security regime could cover all the nuclear security requirements and the implementation of nuclear security measures receive the attention warranted to protect nuclear and radioactive materials, also the existing security systems and measures carry out its purpose in acceptable and satisfactory level. On the safety side public should convince that all safety measures will apply to safe worker, public and environment from any radiological consequences. All that will allow them to trust about their national nuclear security and safety regime and the implementation of its requirements. To enhance the public well understanding towards this critical issue there are some considerations should be taken to turn their apprehension to trust about both national safety and security managements of NSF and HLRW.On the other hand these public apprehensions act as an initiator to the government to develop efficient national policies to manage these critical issues. Also, the study discusses the role of state and organizations to enhance public understanding of the management of nuclear and radioactive materials NSF and HLRW.
This work aims to describe the strategies of management of spent fuel assemblies in the CNAAA power plants adopted as a function of the depletion of the storage capacity of their spent fuel pools. Having chosen the option of dry storage this has the intention to describe the main participants, the Brazilian nuclear policing, the spent fuel storages, the strategy adopted to manage the decrease in storage capacity, the implementation schedule and the current status.
The management of Spent Fuel (SF) is one of the key challenges for nuclear power plant (NPP) operators to tackle, especially in a context of ‘wait-and-see’-policies as generically practiced in many countries. From a scientific-technical perspective, mature options to manage this SF do already exist and can be deployed, being it direct disposal of the SF or reprocessing and subsequent disposal of high-level waste. Though, the socio-political process towards operationalisation of geological disposal is a long one.
SF-management is a long-term activity for utilities and is subject to a variety of uncertain influencing factors as there are:
• Technical-economic: proposed SF-management options differ in industrial maturity as well as having different cost, timing and uncertainty exposures;
• Regulatory: new regulations regarding security of the interim stored SF as well as the geological disposal design and conditions impact the timing and cost of SF-management options;
• Socio-Political: the societal acceptance as well as the (nuclear) energy policy impact the timing and costs of SF-management options;
• Financial: the financial rating of utilities but also the financing mechanisms for the SF-management funds impact the performance of SF-management options.
Utilities therefore face uncertain futures on SF-management translating into financial costs and risks that may increasingly impact the financial rating of particularly utilities with eldering NPPs. Where many studies in the past looked into SF-management options from essentially a technology-push perspective, this paper presents the results of a financial cost/risk-optimising decisioneering methodology on SF-management seeking to optimise utilities’ cost/risk-exposure to uncertain SF-management futures and the risk-mitigating value SF-management options could bring to utilities.
Within such a multi-variate environment, SF-owners do need to seek for lowest cost and increasingly also lowest financial risk strategies covering the next decades. Balancing lower short-term cost options though with higher risks in the longer-term versus higher costs today with lower risks in the future is a decisional challenge and cannot be grasped by assessment methodologies as NPV/DCF.
A combination of stochastic NPV/DCF complemented with Real Options Analysis (ROA) is required to analyse this multi-variate and multi-decades spanning decisional framework. ROA allows to address questions as:
• When to decide which UNF-management option to be executed?
• What’s the value of developments in reducing the cost/risk-exposure for SF-owners in the future?
• What’s the optimal SF-portfolio management for a utilities’ SF and optimised contractual strategy?
The Nuclear Real Options Model (NROM), embedded in the Nuclear Energy System Strategies Assessment Toolbox (NESSAT) by Nuclear-21, is performing such analysis and applied in cost/risk-decisioneering analysis for governments, utilities and fuel cycle service companies.
An example of analysis covering the EU SF-management options is provided in this paper.
The fabrication of EDF's nuclear fuels and their management once used, as well as associated waste, require many industrial operations, qualified as "fuel cycle".
As requested by the French nuclear safety authority (ASN) since 2000, EDF in collaboration with its industrial French partners (Orano Cycle, Framatome, Andra) elaborates periodically a so-called "Impact Cycle" file. This document provides elements to demonstrate for the next ten years the compatibility, in terms of safety and radioprotection, between changes in fuel characteristics and fuel management in NPP and developments in fuel cycle facilities and the corresponding transports.
In June 2016, EDF submitted the file called "Impact cycle 2016" covering the 2016-2030 period.
The IRSN examination focused on:
• the adaptation to needs and evolutions, that may occur in the short or medium term (change of fuel, evolution of facilities, etc.), of the means involved in the fuel cycle (production or storage facilities, logistical means, etc.);
• the study of different scenarios of nuclear-sourced electricity production, including scenarios considering its reduction to 50 % of electricity production by 2025, in accordance with the Energy Transition Law for Green Growth (TECV Law);
• the study of postulated dysfunctions for every stage of the cycle, with identification of associated parades;
• the analysis of major inflections and "cliff effects" that may appear by 2040.
IRSN transmitted to ASN the conclusions of its assessment in May 2018, which were presented to permanent group of experts for laboratories and plants (GPU), including experts from waste, nuclear reactors and transports committees at its meeting on 25 May 2018.
In conclusion, IRSN considers that assessment of the impact on the facilities and the transport activities participating to the French fuel cycle, of the current managements of EDF nuclear fuels and those envisaged until 2030, does not reveal major technical difficulty for this period. The study of prospective scenarios considering a reduction of nuclear-sourced electricity in application of the TECV Law shows that the shutdown of reactors loaded with MOX fuels can induce a short-term saturation of spent nuclear fuel storage facilities. However, a scenario including only the shutdown of reactors loaded with only UOX fuels could delay or even prevent the saturation of these storage facilities. IRSN also underlines the importance of examining the impact on the overall fuel cycle, of reactors shutdown, which will be carried out in application of the TECV Law.
The spent fuel management of nuclear power reactor has two major concerns regarding storage and disposal. Law should comment specifically on the maximum duration of storage of spent fuel and it should ensure that no extended storages will deemed to be de facto disposal. The radioactive waste has potential to extend over several thousand years, it’s always better to have sound spent fuel management schemes so that the future generations won’t be burden by the debts of our generation. Retrieving spent fuel as a further resource will address issues associated with its disposal and security. Majority of laws have future applicability but while drafting nuclear energy/ radioactivity related laws we should provide a clause giving a window for applying technical and safety related advances with retrospective effect to the past activities. Or we should come up with legal principles which are retroactive (retroactive here would like to propose as those incidences which happened in past and still continue to do the same actively) [1]. Due to this new development in the spent fuel management, radioactive waste management will be applicable to the old storage, disposal or mining sites to have new outlook. Regulatory authorities should not only focus on present and future prospects but also required to do reviews of their past operating experiences and lessons learned through their policies. Better risk management is beneficial for shaping public perceptions, supporting spent fuel sites and its acceptance. Improved spent fuel management will not only beneficial for present population but it will also protect the rights of third and fourth generations. This paper also comments on the socio legal requirements across the back end of the fuel cycle. It also comments on necessary improvements in waste management from recycling activities, accident tolerant fuels, and political conflicts among the drivers.
Used fuel is generated from the operation of nuclear reactors of all types. The nuclear industry is currently implementing strategies to ensure the safe and cost-effective management of this used fuel. Currently, there exist two strategies for managing used fuel: the “open cycle” and “closed cycle”. Depending on a number of drivers, countries will engage in one of these alternatives, but may also examine the use of interim storage.
The paper presents used fuel management strategies which ensure used fuel is safely managed by the nuclear operators. The paper also discusses long-term management solutions of used fuel, implementing integrated system approaches in stages to mitigate risks and uncertainties. Additionally, described are innovative solutions that could be implemented in the mid-term, as well as potential constraints to their development.
The “Cycle Impact” approach was launched at the end of the 1990s on the French Nuclear Safety Authority’s initiative (ASN): EDF, in collaboration with its industrial French partners Orano and Andra, has to identify and anticipate actions in order to guarantee a consistent nuclear fuel cycle management in the mid-term. The impacts of fuel design, characteristics and management and nuclear reactors fleet evolutions on the whole supply chain should be studied: NPPs, front-end and back-end facilities, interim storage and logistics. The objective of the “Cycle Impact” exercise is to demonstrate that the choices made by industrial stakeholders on these evolutions, bearing in mind their interdependence and nuclear time frames, do not create unacceptable consequences regarding the entire French fuel cycle. It also allows ASN to have an overview of future regulatory requests to be examined.
After two previous exercises performed in 2000 and 2007, the “Cycle Impact” file 2016, made of 47 deliverables, coordinated by EDF in collaboration with Orano (formerly Areva, French company providing front-end, back-end and logistics services) and Andra (National Agency for Radioactive Waste Management), was submitted in June 2016. The fuel cycle consistency on a 15-years period was analyzed and forecasts until 2040 were made to ensure the absence of cliff-edge effect. Based on four scenarios of nuclear-sourced electricity production and associated fuel reprocessing-recycling strategies parameters like spent fuel pools level of occupancy and transportation casks availability had been studied. Hazard sensitivity analyses had been performed.
On request of the ASN, the file had been assessed by the French technical support organization (IRSN). An expert report including opinions and recommendations was produced and commitments were taken by the industrial actors. The report was subjected to a peer review by the Advisory Committee of experts (GP) for laboratories and plants including experts from waste, nuclear reactors and transports committees. Following the recommendations issued during the meeting of the Advisory Committee on 2018 May 25th, ASN has identified actions to be undertaken by each party in a letter on 2018 October 25th. With the goal of improving nuclear transparency, summary of the IRSN report, GP opinion and ASN stance has been published.
At present the Mining and Chemical Combine is creating an industrial infrastructure of the closed nuclear fuel cycle that is capable of repeated and environmentally safe involvement of reprocessed fuel components in the nuclear fuel cycle.
The spent nuclear fuel (SNF) management infrastructure created presently at the enterprise involves the facilities as follows.
- A water-cooled (wet) storage facility for VVER-1000 spent nuclear fuel;
- An air-cooled (dry) storage facility for RBMK-1000 and VVER-1000 spent nuclear fuel. The dry storage facility is a unique object, being one-of-a kind in the world and using a passive cooling system;
- A start-up facility of the Pilot Demonstration Center for reprocessing spent nuclear fuel with a throughput of 5 spent nuclear fuel tons a year, which was developed on the basis of the best world process engineering solutions for SNF reprocessing;
- A MOX fuel fabrication facility, which is to fabricate fuel to be used in fast reactors and which allows recovered plutonium to be involved in the nuclear fuel cycle.
In addition, for the purpose of SNF transportation, the Combine has a necessary infrastructure and railway rolling stock, which are shipping casks and rail carriages.
The paper encourages not only acknowledging but pro-actively addressing the issues and opportunities resulting from the uncertainty relative to how and when sufficient repository capacity becomes available. It draws on the work of a former IAEA consultancy tasked with addressing very long-term storage of spent nuclear fuel (SNF) as well as current strategic planning and associated initiatives within the U.S. Department of Energy to address the uncertainty in spent fuel management. The paper advocates addressing this uncertainty by design rather than by default. Approaches are suggested for rethinking the basis of spent nuclear fuel storage equipment, facilities, regulatory framework, and communication strategies to acknowledge and proactively address uncertainty relative to the storage duration and the end state of spent nuclear fuel.
As the first phase of the worldwide nuclear fleet is now approaching 40 years of operation, the Back end of the fuel cycle is becoming a forefront focus for utilities having to deal with pool saturation, reactor shutdowns, and requirements for extended periods of interim storage following significant deferral in the implementation of centralized interim storage or geological disposal facilities. As generated radioactive by-products are increasingly being seen as the Achilles heel of our industry, implementation of responsible used fuel management is a condition to ensure sustainability and expansion of nuclear as a low carbon energy source. Given the dynamic and uncertain market environment, cost of electricity and financial performance are not only important to historical utilities but are also key for the development of new capacities in large mature nuclear countries, expanding countries or new comers. In this context, Back end management with its long term liabilities and associated risks has a growing impact on utilities’ financial performance and risk, development potential and market value.
Used fuel and related waste management requires an overarching long-term multi-dimensional system approach which is implemented in stages. A suite of options could be available over the long term, allowing integrating future informed decisions which provide safe, economic solution mitigating risks and uncertainties could be deployed.
Used fuel management system involves multiple decisions over time encompassing conflicts of drivers, uncertain factors and alternatives arising as the market or environmental conditions evolve. Uncertainty and risks are of different natures: technological, environmental, socio- political, economic and financial. Therefore, flexibility in back-end options offers mitigation for the uncertainty of risks. Valuing flexibility and integrating risks when assessing decisions will allow utilities and their stakeholders to decide which option to develop and when.
Orano, providing industrial and innovative back-end solutions and services for over 40 years, will share its developments allowing implementing various alternatives to manage used fuel matching a NPP-operator’s specific financial cost and risk objectives.
National decisions about the management of spent nuclear fuel have global consequences for safety, security, and nonproliferation. For the past four years, the Nuclear Threat Initiative (NTI) has been catalyzing a spent fuel management partnership in the Pacific Rim – East Asia, the United States and Canada – a region with more than 230 power reactors and 170,000+ tons of spent fuel as of 2018. In the last year, the Partnership has created three expert working groups to address specific technical and societal challenges identified by the participants in previous workshops. These are: (1) underground research facility research and development; (2) long-term monitored dry cask storage; and (3) technical and non-technical aspects of repository siting. The working groups will meet several times a year to fulfill their objectives and identify additional topics that would benefit from collaborative research and development.
The paper: surveys the status of nuclear power generation, spent fuel accumulation and spent fuel disposal plans in the Pacific Rim; describes the security and nonproliferation implications of accumulating spent fuel stockpiles; details the efforts that led to the development of a Pacific Rim Spent Fuel Management Partnership and subsequent working groups; and discusses the research agendas of each working group. It is hoped that this Partnership will help provide solutions to practical problems faced by waste managers and can serve as a template for future similar cooperation in other parts of the world.
The U.S. Department of Energy Office of Spent Fuel and Waste Disposition continues to conduct evaluations of removing spent nuclear fuel from shutdown nuclear power plant sites. The evaluations of the 14 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, Vermont Yankee, and Fort Calhoun) provide an important source of information that will be used in future planning for the removal of spent nuclear fuel from the sites. The lessons learned from the evaluations were organized into ten consistent themes. Several of the themes had to do with the shutdown site visits, such as the importance of safety, the importance of preparation for the site visits, the length of the questions submitted to the sites, the importance of using photographs to document site conditions, and the importance of compiling notes at the conclusion of each during a site visit. Several themes had to do with the collection of data, such as the use of Google Earth imagery, the importance of identifying issues associated with the spent nuclear fuel inventory, and the importance of capturing data and experience shortly after shutdown. The remaining themes had to do with building relationships with the shutdown sites and organizations such as Tribes, States, State Regional Group representatives, the Federal Railroad Administration, and local community engagement or advisory panels.
In 2009, Armenian NPP switched to the fuel assemblies with higher initial enrichment that allowed them to reach higher discharge burnups. However, higher burnup that implies bigger decay heat and neutron/gamma irradiation doses requires substantial increase of precooling time in spent fuel pools to meet ANPP NUHOMS type horizontal dry spent fuel storage design acceptance criteria. This may lead to possible issue of availability of enough free cells in spent fuel pools in case of emergency full core unloading. To tackle this issue ANPP decided to estimate additional dose burden during loading and transport on spent fuel in transport casks and storage of them in dry spent fuel storage in case of relatively shorter precooling time satisfying design acceptance criteria on decay heat. For this purpose spent fuel transport cask model was developed by MCNP 6 program. Neutron and gamma source intensities and spectra were calculated by ORIGEN program from SCALE 6 package. Developed model was validated based on irradiation dose measurement results of several spent fuel transport cask loadings.
A sustainable Transportation Used Fuel program could be shortly defined as a not eventful program that you rarely hear about because it is run smoothly with no or minor events. However, used fuel transports are inherently complex due to the nature of the material and the high visibility. It requires a large range of expertise, specialized assets, public acceptance and extremely high performance between the numerous stakeholders to ensure a smooth coordination. As of today, only a few countries have developed comprehensive used fuel transportation program. In the many countries where interim consolidated storage, recycling or geological repository are not available yet, a similar program will have to be implemented.
For more than 50 years Orano TN has safely shipped more than 7,000 used fuel transport casks. The transportation program that was initially developed in the 1970s has been adapted and enhanced over the years to meet more restrictive regulatory requirements and evolving customer needs, and to address public concerns. The numerous “lessons learned” have offered data and guidance that have allowed for also efficient and consistent improvement over the decades...
Based on Orano TN extensive expertise, this paper will describe the different phases and milestones that need to be met to set up, license and operate a successful used fuel transportation management program. Transportation of used fuel in France is performed effectively and efficiently thanks to strong collaboration with the Nuclear Power Plant (NPP) operator who plays a critical role as the shipper.
The performance of spent nuclear fuel (SNF) cladding during transportation and dry storage is an important consideration for demonstrating compliance with the safety-based requirements for transportation and dry storage in the United States. The structural performance of the cladding ensures that the fuel performance remains as analyzed for the transport duration or approved dry storage period. Historically, the U.S. Nuclear Regulatory Commission (NRC) has discussed considerations for ensuring adequate cladding performance through safety review guidance. This guidance has defined adequate fuel conditions, including peak cladding temperatures during short-term loading operations to prevent and mitigate degradation of the cladding. The U.S. NRC has recently supplemented the technical basis in support of the existing guidance on cladding performance by issuance of two draft reports for public comment, NUREG-2224, “Dry Storage and Transportation of High Burnup SNF” and NUREG-2214, “Managing Aging Processes in Storage Report”. This paper will discuss the technical conclusions in these documents and their implications to the regulatory framework for the safety review of high burnup SNF (i.e., SNF with burnups exceeding 45 gigawatt-days/metric ton of uranium). NUREG-2224 addresses the technical issue of hydride reorientation, a process in which the orientation of hydrides precipitated in high burnup SNF cladding during reactor operation changes from the circumferential-axial to the radial-axial direction. NUREG-2224 provides a technical assessment of results from NRC-sponsored research on the effects of hydride reorientation on high burnup SNF cladding performance. NUREG-2224 also provides example approaches for licensing and certification of high burnup SNF for dry storage and transportation, which aim to clarify NRC’s expectations for supporting data on high burnup SNF performance for the evaluation of design-basis drop accidents and vibration normally incident to transport. These approaches recognize the increased flexural rigidity imparted by the fuel pellets on the cladding’s mechanical performance. NUREG-2214 provides a generic evaluation of the age-related mechanisms that have the potential to challenge the ability of SNF cladding to support important-to-safety functions of dry storage systems for periods up to 60 years. These mechanisms consider that time-dependent changes to the cladding are primarily driven by the fuel’s temperature, internal pressure-induced cladding hoop stresses, and the environment during storage or transport operations. Both NUREG-2224 and NUREG-2214 clarify the technical position of the NRC on high burnup SNF performance, which will help improve the effectiveness and efficiency of the review process for applications for dry storage and transportation.
Spain, with 7 operating nuclear power reactors (NPPs) of different technologies and 3 NPPs already shut down, has adopted an open cycle strategy for the back-end of the nuclear fuel cycle. All the spent fuel generated is either wet stored in the spent fuel pools or dry stored in casks from different technologies at the reactor sites. After interim storage, all the spent fuel is expected to be transported to a centralized storage facility that is currently in licensing process but undergoing political controversy. ENSA has specifically developed different technologies of bare fuel type casks for all the Spanish NPPs. Once the initial licensing round of these package designs has been completed and the first casks units are being loaded at some of the NPPs, the next goal is to modify the transportation Certificates of Compliance (CoC) to remove current limitations that restrict the loading of high burnup fuel. Different approaches have been developed and agreed with the nuclear authority depending on the type of fuel rod cladding, the utilisation of the cask and the requirements of the new regulatory standards. Loading of damaged fuel is been approached from two different perspectives. For those power plants with a significant amount of fuel assemblies categorized as ‘damaged’, the ENUN casks will include dedicated basket positions where the entire fuel assembly will be loaded in specific cans. On the opposite, for those NPPs where damaged fuel can be limited to a certain number of fuel rods, ENSA is currently working to adapt specific sealed bundle systems to be loaded in the casks. ENSA’s Engineering area is facing new technical challenges to increase the capabilities of the ENUN cask series to load spent fuel with more demanding requirements, and allow its transportation to the future centralised storage facility
Spent fuel disposal in Finland is in the phase where Posiva is preparing for construction of the encapsulation plant and spent fuel disposal tunnels, and later submitting the operating license application. As part of the preparations, Posiva is planning the future operating phase of the spent fuel encapsulation and final disposal facility. The original operating plan has been to dispose of fuel from both Olkiluoto and Loviisa NPPs in parallel starting around 2024. Another option is to start disposing fuel from Olkiluto NPP first, and continue with Loviisa NPP fuel later. Whatever production option is selected it requires actions and preparations at Loviisa NPP. This paper presents some of these actions.
The storage and secure of SNF should be manage in a safe manner. Egyptian Nuclear Power Plant Authority (ENPPA) is responsible for the safe management of spent nuclear fuel produced by nuclear power plant. The Egyptian nuclear power plant project consists of four Units with VVER-1200 reactors. The operation of 4 NPP units will be accompanied with the accumulation of SNF in the amount of 168 SFAs in average per year. In the course of operation of 4 units during 60 years approximately 11,532 SFAs will be accumulated. TUKs as dual-purpose casks will be used during long-term storage and transportation of the spent fuel assembly for the Egyptian Nuclear Power Plant (ENPP).
The aim of the paper is to establish a proposal for Egyptian safety requirements for safe storage and transportation of spent nuclear fuel. The proposal guided by the Egyptian law on regulation of nuclear activities and radioactivity (law no. 7 of the year 2010) and its Executive regulations, recommendations of IAEA safety standard, the joint convention on the safety of spent fuel management and on the safety of radioactive waste management and Russian nuclear safety requirements. Specific legislation has been established in the nuclear law to: identify responsibility for safety, security and safeguards, and management of spent fuel and radioactive waste (transportation, handling, and storage). Egyptian proposal cover the safety considerations for storage and transportation of spent nuclear fuel.
Short-term operation process of spent fuel means the series of steps occurring in spent fuel building, from loading of spent fuel into shipping casks before transportation. This process includes loading of fuel into shipping casks, sealing of lids, drainage inside canister, drying, and backfilling. The distribution of peak cladding temperature of spent fuel over time tends to be different for wet process, vacuum drying process because heat transfer environments and heat transfer modes are changed. The paper aims to establish thermal analysis scenario on spent fuel short-term operation process for independent verification thermal model.
MCNPX computer code is used to model the general cask GBC-32 which contain 32 typical PWR spent fuel assemblies. For Safe storage and transportation of the cask, factors that affect the criticality were studied using the concept of burn up credit. Several parameters such as initial fuel enrichment, fuel burnup, cooling time, and axial burnup profile were analysed. The analysis was performed in two different steps, first burn the fuel assembly at different burnup and storage conditions, secondly, incorporate the details of the assemblies into the cask and perform a criticality calculations for the cask. Several cases of unnormal storage conditions are considered. The results are compared with similar GBC-32 benchmark.
The paper describes key engineering and logistical solutions for organization of international shipments of nuclear materials from research reactors. The evolution of transport equipment and routes is generalized, and the issues requiring harmonization of national requirements and procedures for the safe transport of the spent nuclear fuel are identified.
Can Spent Nuclear Fuel withstand the shocks and vibrations experienced during normal conditions of transport? This question was the motivation for the multi-modal transportation test conducted in June-October 2017. In this project the US Department of Energy (DOE) (through Sandia National Laboratories and Pacific Northwest National Laboratory) collaborated with the Equipos Nucleares SA, SME (ENSA), Empresa Nacional de Residuos Radiactivos S.A. (ENRESA), and ENUSA Industrias Avanzadas, SA SME (ENUSA) of Spain and Korea Radioactive Waste Agency (KORAD), Korea Atomic Energy Research Institute (KAERI), and Korea Electric Power Corporation Nuclear Fuel (KEPCO NF). The ENsa UNiversal (ENUN) 32P dual-purpose rail cask containing three surrogate PWR assemblies (the assemblies did not contain radioactive fuel) and 29 dummy assemblies (concrete masses) was instrumented with accelerometers and strain gauges. The basket, cask, cradle, and transportation platform were also instrumented. The accelerations and strains were measured during heavy-haul truck, ship, and rail transport, handling operations, and controlled rail tests at the Transportation Technology Center, Inc. (TTCI), a railroad testing and training facility in Pueblo, Colorado. During the test, 40 accelerometers, 37 strain gauges, and three Global Positioning System channels were used to collect 6 terabytes of data over the 54-day, 7-country, 12-state, and 8,500 miles of travel. While strains and accelerations have been measured on the exterior of transportation and storage containers, these measurements have never been collected on the fuel inside the container. The greatest strains and accelerations were observed during the testing at TTCI, specifically during the coupling test. Water transport strains and accelerations were the lowest and heavy haul and rail transport strains and accelerations were comparable. The handling tests were somewhat higher than the most extreme rail tests, except coupling. The observed strains were well below the yield points for spent nuclear fuel cladding demonstrating that the fuel can withstand the shocks and vibrations experienced during normal conditions of transport.
International Nuclear Services (INS) and its majority-owned subsidiary Pacific Nuclear Transport Ltd. (PNTL) have been carrying out transports of different categories of nuclear materials for well over forty years, travelling over five million miles without any nuclear or security incident.
This record has been achieved through significant and sustained investment in the capability of assets, people and systems by INS and its international partners and stakeholders. INS has applied its unique experience in this field to ensure a rigorous and uncompromising approach to regulatory compliance, safety, security, capability and communication.