scholarly journals Exploratory what-if analysis of some debated canister failure modes in the review of a licence application for the construction and operation of a spent nuclear fuel repository in Sweden

2019 ◽  
Vol 49 ◽  
pp. 67-75
Author(s):  
Bo Strömberg ◽  
Lena Sonnerfelt ◽  
Henrik Öberg
Keyword(s):  
Failure Modes ◽  
Spent Nuclear Fuel ◽  
Free Water ◽  
Crystalline Rock ◽  
Spent Fuel ◽  
Creep Ductility ◽  
Regulatory Review ◽  
Bentonite Buffer ◽  
Detailed Assessment ◽  
Long Timescales

Abstract. Regulatory review of the licence application for construction and operation of a spent fuel repository at the Forsmark site in Sweden involves detailed assessment of both expected and hypothetical failure modes of the copper canister. The copper canister, which is supported by the bentonite buffer and the surrounding crystalline rock in the KBS-3 concept, is expected to provide complete containment of radioactive elements for very long timescales. Detailed assessment shows that there is a small probability on such timescales of canister failure due to corrosion following loss of buffer as well as mechanical failure due to large earthquakes. During the regulatory review process, it was proposed that canisters might also fail due to: (i) corrosion in anoxic oxygen gas free water, (ii) pitting corrosion, (iii) stress corrosion cracking, (iv) creep brittle failure, (v) hydrogen embrittlement. We here provisionally accept a number of alternative assumptions related to these processes as a basis for what-if analysis of their implications. The focus is not to determine the merit or to estimate probability of these cases, but rather to explore their potential significance in the context of the available knowledge about the repository environment. Simplified estimates are made of the consequences in terms of number and timing of canister failures as well as radiological impact. It is judged that poor creep ductility of copper would have larger potential consequences compared to localised corrosion phenomena. Potential corrosion failures are expected to be associated with the small fraction of deposition holes that are most extensively exposed to corrodants.

Na/Ca selectivity coefficients of montmorillonite in perchlorate solution at different temperatures

2012 ◽  
Vol 1475 ◽  
Author(s):  
Aku Itälä ◽  
Arto Muurinen
Keyword(s):  
Cation Exchange ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Selectivity Coefficients ◽  
Crystalline Bedrock ◽  
Bentonite Buffer ◽  
Different Temperatures ◽  
C 50 ◽  
Exchange Selectivity

ABSTRACTThe Finnish spent nuclear fuel disposal is based on the Swedish KBS-3 concept in crystalline bedrock. The concept aims at long-term isolation and containment of spent fuel in copper canisters surrounded by bentonite buffer which mostly consists of montmorillonite. For the long-term modelling of the chemical processes in the buffer, the cation-exchange selectivity coefficients have to be known at different temperatures. In this work, the cation-exchange selectivity coefficients and cation-exchange isotherms were determined in batch experiments for montmorillonite at three different temperatures (25 °C, 50 °C and 75 °C). Five different ratios of NaClO4/Ca(ClO4)2 were used in the experimental solutions. After equilibration the solution and montmorillonite were separated and the solution analysed to get the desired exchange parameters. The experiments were modelled with a computational model capable of taking into account the physicochemical processes that take place in the experiment.

scholarly journals Development of a Model for Long-Term Storage Container for VVER-1000 Spent Fuel Assemblies in ANSYS CFX

2017 ◽  
pp. 20-23
Author(s):  
M. Frankova ◽  
Yu. Vorobyov ◽  
M. Vyshemirskiy ◽  
O. Zhabin
Keyword(s):  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Regulatory Review ◽  
Storage Container ◽  
Long Term Storage ◽  
Ansys Cfx ◽  
Calculation Results ◽  
Technical Specifications ◽  
Term Storage

This paper evaluates the thermohydraulic aspects of modeling multi-purpose container (MPC-31) with VVER-1000 spent nuclear fuel (SNF) during its long-term storage in HI-STORM container (designed by Holtec International). The paper describes the main approaches and assumptions applied for the development of correspondent ANSYS CFX model, as well as provides the main calculation results for one of the cases of MPC-31 loading with SNF. The results of calculations were used in the regulatory review of technical specifications of MPC-31 and correspondent safety analysis reports.

Earthquake-induced rock shear through a deposition hole: Laboratory tests on bentonite-material models and modelling of three scale tests

Clay Minerals ◽  
2018 ◽  
Vol 53 (2) ◽  
pp. 213-235
Author(s):  
Lennart Börgesson ◽  
Ann Dueck ◽  
Jan Hernelind
Keyword(s):  
Finite Element ◽  
Laboratory Tests ◽  
Spent Nuclear Fuel ◽  
Material Model ◽  
Spent Fuel ◽  
Stress Strain ◽  
Element Mesh ◽  
Material Models ◽  
Bentonite Buffer ◽  
Nonlinear Stress

ABSTRACTEarthquake-induced rock shear through a bentonite-filled deposition hole in a repository for spent nuclear fuel is an important scenario for the safety analysis because it may cause substantial damage to the canister hosting the spent fuel. Appropriate tools to investigate the effects on the buffer and the canister are required.The study described here explored the laboratory tests conducted to develop a material model of the bentonite buffer to be used in the simulations, the material models that these tests have provided and finite element (FE) simulations of three scale tests of a rock shear for comparison between modelled and measured results. The results were used for validation of the material models and the calculation technique that was used for modelling different rock-shear cases.The laboratory study consisted of swelling-pressure tests and tests to determine shear strength and stress-strain properties. The material model is elastic-plastic with a nonlinear stress-strain relation which depends on the density of the bentonite buffer and is a function of the strain rate. The three scale tests were modelled using theAbaqusfinite element code. Good agreement between modelled and measured results was observed, in spite of the complexity of the models and the difficulties associated with measuring stresses and strains under the very fast shear.The modelling results thus validate the modelling of the SR-Site. The modelling technique, the element mesh and the material models used in these analyses are well fitted and useful for this type of modelling.

Posiva's Site Investigations Programme

2003 ◽  
Vol 807 ◽  
Author(s):  
Veijo Ryhänen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Waste Management ◽  
Nuclear Waste ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Crystalline Rock ◽  
Nuclear Waste Management ◽  
Spent Fuel ◽  
Medium Level

Four nuclear power plant units have been operated in Finland over 20 years. The plants are located at two sites, Olkiluoto and Loviisa. Responsibility for nuclear waste management lies on the utilities, which have established a joint company POSIVA to take care of spent fuel disposal and other expert tasks of nuclear waste management.Already in 1983 the Finnish Government set the objectives and the schedule for the national waste management programme. Since then, two shallow underground repositories have been constructed for low- and medium-level operating waste in crystalline rock at the power plant sites. At the end of 2002 the amount of operating waste emplaced in these two facilities was 4923 m3 (total accumulation 6724 m3).Spent nuclear fuel is stored in interim storage pools at the nuclear power plant sites. The total accumulation was 1228 tU at the end of 2002. Today, the main activities in nuclear waste management concern spent fuel disposal deep in the Finnish bedrock. Apart from technical and scientific issues, a major challenge faced has been the creation of sufficient public acceptance, which is a must in order to obtain favourable political decisions.

The Impact of Alpha-Radiolysis on the Release of Radionuclides from Spent Fuel in a Geologic Repository

1983 ◽  
Vol 26 ◽  
Author(s):  
Ivars Neretnieks
Keyword(s):  
Nuclear Fuel ◽  
Uranium Oxide ◽  
Spent Nuclear Fuel ◽  
Crystalline Rock ◽  
Reducing Agents ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Oxidizing Agents ◽  
Dissolved Iron ◽  
The Impact

ABSTRACTSpent nuclear fuel buried in deep geologic repositories may eventually be wetted by water. The alfa-radiation will radiolyse the water and produce hydrogen and oxidizing agents, mainly hydrogen peroxide and oxygen. The hydrogen will escape by diffusion and the oxidizing agents may attack the canister materials, oxidize the uranium oxide matrix or diffuse out and oxidize reducing agents in the surrounding rock.The rate of radiolysis has been computed recently within the Swedish nuclear fuel safety projects KBS. It is strongly influenced by the amount of available water and by the presence of dissolved iron. The movement of the oxidizing agents out from the canister and their reaction with the reducing agents (mainly ferrous iron) in the Swedish crystalline rock has been modelled as well as the movement of the radionuclides within and past the redox front. Some substances such as uranium, neptunium and technetium will precipitate at the redox front and will be withdrawn from the water to a considerable extent.

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.

Some Important Mechanisms and Processes in the Near Field of the Swedish Repository for Spent Nuclear Fuel

1992 ◽  
Vol 294 ◽  
Author(s):  
Ivars Neretnieks
Keyword(s):  
Molecular Diffusion ◽  
Spent Nuclear Fuel ◽  
Dominant Role ◽  
Near Field ◽  
Water Flow Rate ◽  
Crystalline Rock ◽  
Diffusive Transport ◽  
Actual Design ◽  
Natural Analogues ◽  
And Performance

ABSTRACTIn repositories for nuclear waste there are many processes that will be instrumental in damaging the canisters and releasing the nuclides. Based on experiences from studies of the performance of repositories and of an actual design, the major mechanisms influencing the integrity and performance of a repository are described and discussed. The paper addresses only conditions in crystalline rock repositories. The low water flow rate in fractures and channels plays a dominant role in limiting the interaction between water and waste. Molecular diffusion in the backfill and rock matrix, as well as in the mobile water, is an important transport process, but actually limits the exchange rate because diffusive transport is slow. Solubility limits of both waste matrix and of individual nuclides are also important. Complicating processes include alpha-radiolysis, which may change the water chemistry in the near-field. The sizes and locations of water flowpaths and damages in the canisters considerably influence the release rates. Uncertainties in data are large. Nevertheless the system is very robust in the sense that practically no reasonably conceivable assumptions or data will lead to large nuclide releases. Several natural analogues have been found to exhibit similarities with a waste repository and help to validate concepts and to increase our confidence that all major issues have been considered.

Safety Assessment of Long Term Serviced Pressure Vessels: A Case Study of Typical Refining and Chemical Plants in China

2021 ◽  
Author(s):  
Zhiyuan Han ◽  
Guoshan Xie ◽  
Haiyi Jiang ◽  
Xiaowei Li
Keyword(s):  
Pressure Vessel ◽  
Safety Assessment ◽  
Failure Modes ◽  
Pressure Vessels ◽  
Time Dependent ◽  
Chemical Plants ◽  
Safety Specification ◽  
Detailed Assessment

Abstract The safety and risk of the long term serviced pressure vessels, especially which serviced more than 20 years, has become one of the most concerned issues in refining and chemical industry and government safety supervision in China. According to the Chinese pressure vessel safety specification TSG 21-2016 “Supervision Regulation on Safety Technology for Stationary Pressure Vessel”, if necessary, safety assessment should be performed for the pressure vessel which reaches the design service life or exceeds 20 years without a definite design life. However, the safety and risk conditions of most pressure vessels have little changes after long term serviced because their failure modes are time-independent. Thus the key problem is to identify the devices with the time-dependent failure modes and assess them based on the failure modes. This study provided a case study on 16 typical refining and chemical plants including 1870 pressure vessels serviced more than 20 years. The quantitative risk and damage mechanisms were calculated based on API 581, the time-dependent and time-independent failure modes were identified, and the typical pressure vessels were assessed based on API 579. Taking the high pressure hydrogenation plant as an example, this study gave the detailed assessment results and conclusions. The results and suggestions in this study are essential for the safety supervision and extending life of long term serviced pressure vessels in China.

Digital autoradiography on C-14-labelled PMMA impregnated rock samples using the BeaverTM

MRS Advances ◽  
2018 ◽  
Vol 3 (21) ◽  
pp. 1161-1166 ◽  
Author(s):  
Mikko Voutilainen ◽  
Juuso Sammaljärvi ◽  
Eveliina Muuri ◽  
Jérôme Donnard ◽  
Samuel Duval ◽  
...  
Keyword(s):  
Spatial Distribution ◽  
Spent Nuclear Fuel ◽  
Rock Sample ◽  
Crystalline Rock ◽  
Imaging Plate ◽  
Heterogeneous Distribution ◽  
Heterogeneous Structures ◽  
In Situ Diffusion

In Finland and Sweden the KBS-3 concept has been chosen for the disposal of spent nuclear fuel in crystalline rock. Recent experiments have shown that heterogeneity of rock may play a major role in the transport of radionuclides. Autoradiographic methods have been proven to be able to assist the characterization of heterogeneous structures. In this study we tested a novel filmless autoradiographic device called BeaverTM which applies a micro patterned gaseous detector in order to quantitatively map beta emissions from C-14 atoms. The studied samples were impregnated with C-14-labelled methylmethacrylate (C-14-MMA) and polymerized to C-14-PMMA with thermal initiator. The BeaverTM was then used to determine the spatial distribution of the C-14-PMMA by measuring the C-14 emissions. The porosity is determined from the amount of C-14-PMMA in the rock sample and results were compared to ones from phosphor imaging plate autoradiography. The resulting images show a heterogeneous distribution of porosity which arises from the different minerals. The samples were chosen from three sites that have been used recently for in situ diffusion experiments: Olkiluoto (Finland), Äspö (Sweden) and Grimsel (Switzerland).

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