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.

A New Approach to the RN Source Term for Spent Nuclear Fuel under Geological Disposal Conditions

2000 ◽  
Vol 663 ◽  
Author(s):  
Christophe Poinssot ◽  
Christophe Jegou ◽  
Pierre Toulhoat ◽  
Jean-Paul Piron ◽  
Jean-Marie Gras
Keyword(s):  
Low Temperature ◽  
Closed System ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Driving Forces ◽  
Spent Fuel ◽  
Microscopic Description ◽  
New Approach ◽  
Geological Disposal ◽  
Source Term Model

ABSTRACTUnder the geological disposal conditions, spent fuel (SF) is expected to evolve during the 10,000 years while being maintained isolated from the biosphere before coming in contact with water. Under these circumstances, several driving forces would lead to the progressive intrinsic transformations within the rod which would modify the subsequent release of radionuclides. The major mechanisms are the production of a significant volume of He within the UO2 lattice, the accumulation of irradiation defects due to the low temperature which avoids any annealing, the slow migration of radionuclides (RN) within the system by (i) the α self-irradiation-induced athermal diffusion and (ii) locally the building-up of internal mechanical stresses which could turn the pellets into powder. However, the current RN source terms for SF have never accounted for this evolution as they are based on existing knowledge of 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 take account of the early evolution of the closed system, the amplitude of which increases with the burnup and is greater for MOX fuels.

scholarly journals An Automated SeaFAST ICP-DRC-MS Method for the Determination of 90Sr in Spent Nuclear Fuel Leachates

Molecules ◽  
2020 ◽  
Vol 25 (6) ◽  
pp. 1429 ◽  
Author(s):  
Víctor Vicente Vilas ◽  
Sylvain Millet ◽  
Miguel Sandow ◽  
Luis Iglesias Pérez ◽  
Daniel Serrano-Purroy ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Analytical Approach ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Chemical Purification ◽  
Radiochemical Method ◽  
Good Agreement

To reduce uncertainties in determining the source term and evolving condition of spent nuclear fuel is fundamental to the safety assessment. ß-emitting nuclides pose a challenging task for reliable, quantitative determination because both radiometric and mass spectrometric methodologies require prior chemical purification for the removal of interfering activity and isobars, respectively. A method for the determination of 90Sr at trace levels in nuclear spent fuel leachate samples without sophisticated and time-consuming procedures has been established. The analytical approach uses a commercially available automated pre-concentration device (SeaFAST) coupled to an ICP-DRC-MS. The method shows good performances with regard to reproducibility, precision, and LOD reducing the total time of analysis for each sample to 12.5 min. The comparison between the developed method and the classical radiochemical method shows a good agreement when taking into account the associated uncertainties.

Assessment of the Evolution with Time of the Instant Release Fraction of Spent Nuclear Fuel in Geological Disposal Conditions

2002 ◽  
Vol 713 ◽  
Author(s):  
Christophe Poinssot ◽  
Patrick Lovera ◽  
Marie-Hélène Faure
Keyword(s):  
Nuclear Fuel ◽  
Quantitative Assessment ◽  
Spent Nuclear Fuel ◽  
Diffusion Processes ◽  
Driving Forces ◽  
Geological Disposal ◽  
Physical And Chemical ◽  
Evolution With Time ◽  
New Framework

ABSTRACTUnder the geological disposal conditions, spent nuclear fuel (SNF) is expected to evolve during the first thousands years while being maintained isolated from the biosphere before water comes in. Under those circumstances, several driving forces would lead to the progressive intrinsic SNF transformations within the rod which would basically modify the physical and chemical state of the fuel and the subsequent release of radionuclides in solution. In this paper, we briefly summarize the mechanisms we estimate to be significant and propose a new framework for the quantitative assessment of the radionuclide (RN) inventory we estimate to be associated to the classically referred to “Instant Release Fraction” (IRF). We hence demonstrate that in this framework, significantly high IRF values have to be expected for the long term due mainly to the presence of athermal diffusion processes.

Review about the Effect of He on the Microstructure of Spent Nuclear Fuel in a Repository

MRS Advances ◽  
2016 ◽  
Vol 1 (62) ◽  
pp. 4147-4156 ◽  
Author(s):  
C. Ferry ◽  
J. Radwan ◽  
H. Palancher
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Critical Concentration ◽  
Fuel Pellet ◽  
Spent Fuel ◽  
Operational Model ◽  
Matrix Microstructure ◽  
Natural Analogues ◽  
Order Of Magnitude ◽  
Radionuclide Release

ABSTRACTHelium is produced in spent nuclear fuel by α-decays of actinides. After 10,000 years, the concentration of He accumulated in UO2 spent fuel is about 0.23 at.%. For direct disposal of spent nuclear fuel, consequences of helium build-up on the fuel matrix microstructure must be evaluated since it can modify the radionuclide release when water comes into contact with the spent fuel surface, after breaching of the disposal canister. An operational model has been proposed in order to evaluate the effect of helium on the microstructure of spent fuel in a repository. Based on conservative assumptions and different scenarios of bubble population, the calculated helium critical concentration, that could lead to a partial loss of integrity of the spent fuel pellet, is 0.37 at.%. However, observations on He-implanted UO2, α-doped UO2 pellets and natural analogues evidence a macroscopic damage only for He concentrations, which are more than one order of magnitude higher.

Release of 108Ag from Irradiated PWR Control Rod Absorbers under Deep Repository Conditions

2015 ◽  
Vol 1744 ◽  
pp. 217-222
Author(s):  
O. Roth ◽  
M. Granfors ◽  
A. Puranen ◽  
K. Spahiu
Keyword(s):  
Nuclear Fuel ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Nuclear Reactors ◽  
Reactor Irradiation ◽  
Spent Fuel ◽  
Control Rod ◽  
Potential Source ◽  
Absorber Material ◽  
Control Rods

ABSTRACTIn a future Swedish deep repository for spent nuclear fuel, irradiated control rods from PWR nuclear reactors are planned to be stored together with the spent fuel. The control rod absorber consists of an 80% Ag, 5% Cd, 15% In alloy with a steel cladding. Upon in-reactor irradiation 108Ag is produced by neutron capture. Release of 108Ag has been identified as a potential source term for release of radioactive substances from the deep repository.Under reducing deep repository conditions, the Ag corrosion rate is however expected to be low which would imply that the release rate of 108Ag should be low under these conditions. The aim of this study is to investigate the dissolution of PWR control rod absorber material under conditions relevant to a future deep repository for spent nuclear fuel. The experiments include tests using irradiated control rod absorber material from Ringhals 2, Sweden. Furthermore, un-irradiated control rod absorber alloy has been tested for comparison. The experiments indicate that the release of Ag from the alloy when exposed to water is strongly dependent on the redox conditions. Under aerated conditions Ag is released at a significant rate whereas no release could be measured after 133 days during leaching under H2.

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.

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.

The Assessment of the Long-Term Evolution of the Spent Nuclear Fuel Matrix by Kinetic, Thermodynamic and Spectroscopic Studies of Uranium Minerals.

1994 ◽  
Vol 353 ◽  
Author(s):  
Jordi Bruno ◽  
I. Casas ◽  
E Cera ◽  
R. C. Ewing ◽  
R. J. Finch ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Spectroscopic Studies ◽  
Spent Fuel ◽  
Term Evolution ◽  
Uranium Minerals ◽  
The Stability

AbstractThe long term behaviour of spent nuclear fuel is discussed in the light of recent thermodynamic and kinetic data on mineralogical analogues related to the key phases in the oxidative alteration of uraninite. The implications for the safety assessment of a repository of the established oxidative alteration sequence of the spent fuel matrix are illustrated with Pagoda calculations. The application to the kinetic and thermodynamic data to source term calculations indicates that the appearance and duration of the U(VI) oxyhydroxide transient is critical for the stability of the fuel matrix.

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.

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