scholarly journals International safeguards for the final disposal of spent nuclear fuel – why, what and how

2021 ◽  
Vol 1 ◽  
pp. 241-242
Author(s):  
Irmgard Niemeyer ◽  
Katharina Aymanns ◽  
Guido Deissmann ◽  
Dirk Bosbach
Keyword(s):  
Member State ◽  
Nuclear Fuel ◽  
Information Needs ◽  
Spent Nuclear Fuel ◽  
Nuclear Material ◽  
Specific Information ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Final Disposal ◽  
International Safeguards

Abstract. The objectives of international safeguards are the timely detection of diversion of significant quantities of nuclear material from peaceful nuclear activities to the manufacture of nuclear weapons (or for other purposes), and deterrence of such diversion by the risk of early detection for states with comprehensive safeguards agreements with the International Atomic Energy Agency (IAEA). Following these objectives, several studies have focused on the developments of concepts and methods for safeguarding final disposal facilities as well as on identifying the most feasible technologies that could potentially be deployed for verifying final disposal programmes (IAEA, 1998, 2010, 2018). These activities were coordinated through Member State Safeguards Support Programmes, including the joint tasks on the development of “Safeguards for Geological Repositories” (SAGOR, 1994–2004) and on the “Application of Safeguards to Geological Repositories” (ASTOR, 2005–2017). SAGOR performed a diversion path analysis for spent fuel disposal facilities, determined safeguards technical objectives and identified potential safeguards measures to meet those objectives. ASTOR supported the IAEA in assessing how safeguards measures could be effectively implemented and provided recommendations with respect to developing such measures. Specific verification technologies were developed under other Member State Support Programme tasks. A summary report on the progress and status of safeguards for spent fuel encapsulation plants and geological repositories was completed by ASTOR in 2017. ASTOR also identified areas and actions that need to be accomplished to support safeguards implementation in final disposal facilities, such as (1) establish performance requirements for the design of safeguards technologies relevant to geological disposal of spent fuel, (2) determine specific information needs of states and operators regarding safeguards implementation for geological disposal of spent fuel and develop appropriate guidance, (3) determine specific information needs of IAEA inspectors and analysts and develop a guidance document that provides recommendations for implementing safeguards for a geological repository system under the state-level concept and (4) develop and test appropriate safeguards equipment (IAEA, 2017; Moran et al., 2018). While several measures and technologies related to verifying the geological disposal of spent fuel have been used by the IAEA at other facilities or are in development or testing, other technologies still need to be developed and tested. In addition, ASTOR identified the need for approaches to how information about disposed spent fuel and high-level nuclear waste should be managed, handled, organized, archived, read, interpreted and secured for the long term (for centuries after repository closure and beyond), including an international standard for states and facility operators on information management, data-retention methods and timescales for preserving safeguards data for geological repositories. The presentation will introduce the objectives of international nuclear material safeguards for the final disposal of spent nuclear fuel, highlight the current status of developments and discussions in terms of approaches and technologies for safeguarding geological repositories, and give an outlook on implementing safeguards for final disposal in Germany.

Preparing the Spent Nuclear Fuel for Long-Term Storage and Final Disposal by Encapsulating It Into Glass-Ceramic Waste Forms

2001 ◽  
Author(s):  
Sergey Yu. Sayenko ◽  
G. A. Kholomeyev ◽  
B. A. Shilyaev ◽  
A. V. Pilipenko ◽  
E. P. Shevyakova ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Glass Ceramic ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Final Disposal ◽  
Ceramic Composition ◽  
Long Term Storage ◽  
Term Storage

Abstract This paper describes the research work carried out at the NSC KIPT to develop and apply a final waste form in the shape of a monolithic solid block for the containment of spent nuclear fuel. To prepare radioactive waste for long-term storage and final deep geological disposal, investigations into the development of methods of immobilizing HLW simulators in protective solid matrices are being conducted at the NSC KIPT. For RBMK spent nuclear fuel it is proposed and justified to encapsulate the spent fuel bundles into monolithic protective blocks, produced with the help of hot isostatic pressing (HIP) of powder materials. In accordance with this approach, as a material for the protective block made up of the glass-ceramic composition prepared by sintering at isostatic pressure, the powder mixture of such natural rocks as granite and clay has been chosen. Concept approach and characterization of waste form, technological operations of manufacturing and performance assessment are presented. The container with spent fuel for long-term storage and final disposal presents a three barrier protective system: ceramic fuel UO2 in cladding tube, material of the glass-ceramic block, material of the sealed metal capsule. Investigations showed that the produced glass-ceramic material is characterized by high stability of chemical and phase compositions, high resistance in water medium, low porosity (compared with the porosity of natural basalt). With the help of mathematical calculations it was shown that the absorbed dose of immobilizing material by RBMK spent fuel irradiation for 1000 years of storage in the geological disposal after 10 years of preliminary cooling will be ∼ 3.108 Gy, that is 2–3 orders of magnitude less than the values corresponding to preserving radiation resistance and functional parameters of glasses and ceramics. The average value of velocity of linear corrosion in water medium of the protective layer made up of the glass-ceramic composition determined experimentally makes up ∼ 15 mm per year. This allows to use glass-ceramic compositions effectively as an engineering barrier in the system of spent fuel geological disposal and to increase the lifetime of the waste container, in particular, up to 3000 years with the layer thickness ∼ 40 mm. The possible release of radionuclides from the waste container during its interim storage in the open air (near-surface storage) is estimated. The calculations are made by taking into account the possible increase of coefficients of radionuclide diffusion from 10−16 to 10−14 m2/c as a result of spent fuel radiation affecting the protective layer. The obtained results showed that the protective barrier (about 40 mm) at the base of the glass-ceramic composition, ensures reliable isolation from the environment against the release of radionuclides from the controlled near-surface long-term storage far up to 1000 years. The relatively limited release of radionuclides will make up about 1% for the period of more than 400 years, and 10% - in 1000 years. During this period of time, the radionuclides 90Sr and 137Cs will completely turn into stable 90Zr and 137Ba and the decay of many transuranium elements will occur. The results from laboratory scale experiments, tests and calculations carried out so far, show that the proposed glass-ceramic materials may be used as basic materials for manufacturing the monolithic protective block in which the spent fuel elements will be embedded with the aim of further long-term storage or final disposal.

Long Term Storage of Nuclear Spent Fuel as Key Role of Japan’s Nuclear Fuel Cycle Until 2100: Cost and Benefit

2010 ◽  
Author(s):  
Tadahiro Katsuta
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
Storage Site ◽  
Spent Fuel ◽  
Final Disposal ◽  
Enriched Uranium ◽  
Interim Storage ◽  
Reprocessing Plant

Political and technical advantages to introduce spent nuclear fuel interim storage into Japan’s nuclear fuel cycle are examined. Once Rokkasho reprocessing plant starts operation, 80,000 tHM of spent Low Enriched Uranium (LEU) fuel must be stored in an Away From Reactor (AFR) interim storage site until 2100. If a succeeding reprocessing plant starts operating, the spent LEU will reach its peak of 30,000 tHM before 2050, and then will decrease until the end of the second reprocessing plant operation. Throughput of the second reprocessing plant is assumed as twice of that of Rokassho reprocessing plant, indeed 1,600tHM/year. On the other hand, tripled number of final disposal sites for High Level Nuclear Waste (HLW) will be necessary with this condition. Besides, large amount of plutonium surplus will occur, even if First Breeder Reactors (FBR)s consume the plutonium. At maximum, plutonium surplus will reach almost 500 tons. These results indicate that current nuclear policy does not solve the spent fuel problems but rather complicates them. Thus, reprocessing policy could put off the problems in spent fuel interim storage capacity and other issues could appear such as difficulties in large amount of HLW final disposal management or separated plutonium management. If there is no reprocessing or MOX use, the amount of spent fuel will reach over 115,000 tones at the year of 2100. However, the spent fuel management could be simplified and also the cost and the security would be improved by using an interim storage primarily.

Radionuclides Release Model for Performance Assessment Studies of Spent Nuclear Fuel in Geological Disposal

2004 ◽  
Vol 824 ◽  
Author(s):  
Christophe Poinssot ◽  
Patrick Lovera ◽  
Cécile Ferry
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Hydrogen Effect ◽  
Spent Fuel ◽  
Geological Disposal ◽  
The Matrix ◽  
Simple Kinetic Model ◽  
Scientific Results ◽  
Release Model

AbstractIn the framework of the research conducted on the long term evolution of spent nuclear fuel in geological disposal conditions, a source term model has been developed to evaluate the instantaneous release of RN (Instant Release Fraction IRF) and the delayed release of the RN which are embedded within the matrix. This model takes into account all the scientific results currently available in the literature except the hydrogen effect. IRF was assessed by considering the evolution with time of the RN inventories located within the fuel microstructure to which no confinement properties can be allocated on the long term (rim, gap, grain boundaries). It allows to propose some reference bounding values for the IRF as a function of time of canister breaching and burnup. The matrix radiolytic dissolution was modeled by a simple kinetic model neglecting the radiolytic species recombination and the influence of aqueous ligands and radiolytic oxidants were supposed to completely react with the fuel surface. Spent fuel performance was therefore demonstrated to deeply depend on the reactive surface area.

Operational Model for the Spent Nuclear Fuel Radionuclides Source Term in Presence of Water

2001 ◽  
Author(s):  
Christophe Poinssot ◽  
Christophe Jegou ◽  
Pierre Toulhoat ◽  
Jean-Marie Gras
Keyword(s):  
Nuclear Fuel ◽  
Closed System ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Driving Forces ◽  
Spent Fuel ◽  
Microscopic Description ◽  
Operational Model ◽  
Geological Disposal ◽  
Source Term Model

Abstract Under the geological disposal conditions, spent fuel (SF) is expected to evolve during the 10,000 years while being maintained isolated from the biosphere before water comes in. Under those circumstances, several driving forces would lead to the progressive intrinsic transformations within the rod which would modify the subsequent release of radionuclides: the production of a significant volume of He, the accumulation of irradiation defects, the slow migration of radionuclides (RN) within the pellet. However, the current RN source terms for SF never accounted for these evolutions and was based on the existing knowledge on the fresh SF. Two major mechanisms were considered, the leaching of the readily available fraction (one which was supposed to be instantly accessible to water), and the release of RN through alteration of the UO2 grains. We are now proposing a new RN source term model based on a microscopic description of the system in order to also account for the early evolution of the closed system, the amplitude of which increases with the burnup and is greater for MOX fuels.

UK Contributions to the Decommissioning of the BN-350 Reactor in Kazakhstan: 2002–2011

2011 ◽  
Author(s):  
Steven Laws ◽  
David Wells ◽  
Andrew Herrick
Keyword(s):  
Liquid Metal ◽  
Nuclear Fuel ◽  
Fast Reactor ◽  
Spent Nuclear Fuel ◽  
Nuclear Material ◽  
Department Of Energy ◽  
Spent Fuel ◽  
Highly Active ◽  
The Usa ◽  
Storage Pond

Since 2002, the UK’s Global Threat Reduction Programme managed by the Department of Energy and Climate Change (DECC) has provided assistance to the Republic of Kazakhstan with the decommissioning of the BN-350 sodium cooled fast reactor. Assistance has focused on non-proliferation, safety and security projects to ensure the permanent and irreversible shutdown of the reactor and the reduction of security, safety and environmental hazards, particularly those associated with the large inventory of liquid metal coolants (sodium and sodium-potassium alloy) and the presence of spent nuclear fuel (SNF). UK assistance efforts have been co-ordinated with those of the USA and have made use of the UK’s experience in decommissioning its own fast reactor plants at Dounreay. The paper describes work undertaken with UK technical and funding assistance support in the following areas: • Provision of training and technical support in project management and technical topics, including assistance with completion of the BN-350 Decommissioning Plan. • Liquid metal coolant treatment projects, including immobilization of liquid products from the Sodium Processing Facility (SPF) and processing of residual sodium remaining within the drained coolant circuits. • Immobilization of highly active caesium traps, arising from sodium clean-up both during reactor operations and post-shutdown. • Operations to transfer the entire inventory of spent nuclear fuel from the reactor storage pond into dual-use storage and transport casks and then consign these casks to long-term secure storage remote from the reactor site. This activity was part of the major US-Kazakhstan SNF Storage Project. • Surveys of spent fuel route facilities to establish the absence of any significant amount of nuclear material. Key achievements in 2010 were the successful completion of residual sodium processing and completion of the SNF Storage Project. Through 2011, it is intended that the surveys of the fuel storage pond and immobilization of caesium traps will be completed, bringing the current UK assistance activities to an end before March 2012.

Key Scientific Issues Related to the Evolution of Spent Nuclear Fuel in Long Term Dry Storage and Geological Disposal

2003 ◽  
Vol 807 ◽  
Author(s):  
Christophe POINSSOT ◽  
Cécile FERRY ◽  
Jean-Marie GRAS
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Various Boundary Conditions ◽  
Dry Storage ◽  
Mechanical Evolution ◽  
Key Issues ◽  
Physical And Chemical

ABSTRACTThe anticipated long term evolution of spent nuclear fuel as well as the remaining scientific key issues are presented for the various boundary conditions that can be encountered in long term dry storage and geological disposal. Spent fuel is expected to evolve significantly in closed system conditions which are representative of long term dry storage and the first stages of geological disposal. The mechanical evolution of the grain boundaries, the fate of helium and the evolution of the RN location within the pellet are the three major questions to be addressed which could significantly modify the physical and chemical state of the fuel. In addition, mechanisms and kinetics of fuel alteration by water in deep geological repository are still to be more deeply understood, in particular the inventory of the instant release and the radiolytic dissolution processes, to get a robust and reliable source term.

Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel

2006 ◽  
Vol 985 ◽  
Author(s):  
Jeffrey A. Fortner ◽  
A. Jeremy Kropf ◽  
James L. Jerden ◽  
James C. Cunnane
Keyword(s):  
Transition Zone ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Aqueous Environment ◽  
Spent Fuel ◽  
Aqueous Corrosion ◽  
Reduction Potentials ◽  
Chemical Effects ◽  
Before And After ◽  
X Ray Absorption

AbstractPerformance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO2 matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO2 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO22+) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of ∼ 50 micrometers. We find evidence of a thin (∼ 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO2+/Np4+ and UO22+/U4+ couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions.

scholarly journals The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

MRS Advances ◽  
2018 ◽  
Vol 3 (19) ◽  
pp. 991-1003 ◽  
Author(s):  
Evaristo J. Bonano ◽  
Elena A. Kalinina ◽  
Peter N. Swift
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
United States Of America ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
The United States ◽  
Spent Fuel ◽  
Dry Storage ◽  
The Us

ABSTRACTCurrent practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-century when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.

scholarly journals Nuclear Data Uncertainty Quantification in Criticality Safety Evaluations for Spent Nuclear Fuel Geological Disposal

2021 ◽  
Vol 11 (14) ◽  
pp. 6499
Author(s):  
Matthias Frankl ◽  
Mathieu Hursin ◽  
Dimitri Rochman ◽  
Alexander Vasiliev ◽  
Hakim Ferroukhi
Keyword(s):  
Uncertainty Quantification ◽  
Nuclear Fuel ◽  
Evaluation Method ◽  
Spent Nuclear Fuel ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Evaluation Methodology ◽  
Geological Disposal ◽  
Fuel Assemblies ◽  
Criticality Safety

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO2 fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies.

Final disposal of spent nuclear fuel in Taiwan: A state-of-the-art technical overview

1989 ◽  
Vol 9 (3) ◽  
pp. 171-188 ◽  
Author(s):  
Shang-Jyh Liu ◽  
Soong Kuo-Liang ◽  
Yang Jing-Tong
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
State Of The Art ◽  
Final Disposal
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