Modeling and simulation of ductile damage in forged steel used to encapsulate high-level radioactive waste

2021 ◽  
pp. 105678952110617
Author(s):  
Jérémy Serveaux ◽  
Carl Labergere ◽  
Frédéric Bumbieler ◽  
Khémais Saanouni
Keyword(s):  
Radioactive Waste ◽  
Image Correlation ◽  
Ductile Damage ◽  
Loading Path ◽  
Irreversible Processes ◽  
High Level Waste ◽  
Cauchy Stress ◽  
Notched Specimens ◽  
High Level

Andra, the French national radioactive waste management agency, is in charge of studying the disposal of high-level and long-lived intermediate-level waste (HLW and ILW-LL) in a deep geological repository. According to the reference concept, it is planned to encapsulate high-level waste in non-alloy P285NH steel overpacks before inserting them into horizontal steel cased micro-tunnels. This work is a part of the study about the long-term behavior of a welded steel overpack subjected to external hydrostatic pressure and several localized loading paths. Indeed, the main objective of this work is to develop the most suitable model for non-alloy steel P285NH to be used in the prediction of the long-term overpack behavior. Dealing with a ductile steel, elastoplastic constitutive equations accounting for mixed nonlinear isotropic and kinematic hardening strongly coupled with ductile isotropic damage are adopted. They are formulated based on the classical thermodynamics of irreversible processes framework with state variables at the macroscopic scale, (Germain, 1973) (Lemaitre 1985, Saanouni 2012). In this paper, a new coupling formulation between the scalar isotropic ductile damage and the deviatoric and spherical part of the Cauchy stress and elastic strain tensors is proposed. In order to calibrate the developed model on P285NH steel, multiple tensile tests are performed using classical cylindrical specimens, notched specimens and double notched specimens. In the last part, some experimental fields are measured using digital image correlation. Application is made to a simplified overpack represented by thick walled cylinder subject to compressive loading path. A FEM (Finite Element method) crushing operation of an overpack’s cylindrical part has simulated and analysed.

Issues in the Timing of High Level Waste Disposal: An International Perspective

2008 ◽  
Author(s):  
Stan Gordelier ◽  
Pa´l Kova´cs
Keyword(s):  
Radioactive Waste ◽  
Waste Disposal ◽  
Nuclear Power ◽  
Nuclear Industry ◽  
High Level Waste ◽  
Geological Disposal ◽  
The Public ◽  
Deep Geological Disposal ◽  
High Level

The world is facing energy difficulties for the future, in terms of security of supply and climate change issues. Nuclear power is virtually carbon free and it contributes to energy security, being a quasi-domestic source. Whilst it cannot provide a complete answer to these challenges, it is certainly capable of providing a significant component of the answer. However, nuclear power remains controversial. In order to gain public acceptance, it is widely recognised that a number of key issues need to be addressed, amongst which is resolution of the high-level radioactive waste (HLW) (including spent fuel) disposal issue. This is an important issue for all countries with an existing nuclear programme, whether or not it is intended that nuclear power should be phased out or expanded — the waste already exists and must be managed in any event. It is equally important for countries planning a new nuclear power programme where none has previously existed. Since nuclear power was first developed over fifty years ago, HLW arisings have been stored as an interim measure. It is widely believed by experts (though not by many opponents of the nuclear industry, nor by the public) that deep geological disposal, after a reasonable cooling time in interim storage, is technically feasible and constitutes a safe option [1] at an acceptable cost. The total volume of HLW from nuclear reactors is relatively small. A key issue, however, is the time-scale for developing such a final disposal solution. Considerations of security and inter-generational equity suggest that geological disposal should be implemented as soon as possible irrespective of whether or not new arisings are created. The question of managing HLW is not necessarily related to the issue of building new nuclear power stations. However, many opponents argue that there has been insufficient demonstration of the long-term safety of deep geological disposal. The same opponents also argue that there should be a moratorium on building new nuclear power plants (NPPs) until the issue of long-term management of HLW is resolved. These arguments have a powerful influence on public opinion towards both the construction of a waste repository and the building of new NPPs. The intent of this paper (developed from the current OECD NEA study on “Timing of High Level Waste Disposal”) is to identify and discuss some of the factors influencing the timing of the implementation of a HLW disposal strategy and to demonstrate to decision makers how these factors are affecting country strategies, based on current experience. Determining an optimum timescale of HLW disposal may be affected by a wide range of factors. The study examines how social acceptability, technical soundness, environmental responsibility and economic feasibility impact on the timing of HLW disposal and can be balanced in a national radioactive waste management strategy taking the social, political and economic environment into account. There is clear evidence that significant fractions of the public still have serious misconceptions with respect to the issues surrounding nuclear waste. The nuclear industry, together with governments in those countries who would like a component of nuclear power in their energy mix, has a responsibility for and a significant challenge in presenting its case to the public.

The Belgian R&D Feasibility Programme on the Geological Disposal of High-Level and Long-Lived Radioactive Waste

2013 ◽  
Author(s):  
Philippe Van Marcke ◽  
William Wacquier
Keyword(s):  
Radioactive Waste ◽  
High Level Waste ◽  
Reference Solution ◽  
Conceptual Level ◽  
Geological Disposal ◽  
Low Level ◽  
High Level ◽  
Term Management

ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, considers geological disposal in poorly indurated clay as the reference solution for the long-term management of high-level waste (HLW) and intermediate and low level waste, long-lived (ILLW-LL). The disposal concept entails the post-conditioning of the waste in disposal packages and the subsequent disposal of these packages in an underground repository. The R&D feasibility programme on geological disposal aims at demonstrating, at a conceptual level, that the proposed disposal system can be constructed, operated and closed.

Safety and Sustainability Implications of Long Term Storage of Radioactive Waste

2003 ◽  
Author(s):  
John Rowat
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Information Transfer ◽  
Spent Nuclear Fuel ◽  
Position Paper ◽  
High Level Waste ◽  
Long Term Storage ◽  
High Level ◽  
Term Storage

Storage and disposal of radioactive waste are complementary rather than competing activities, and both are required for the safe management of wastes. Storage has been carried out safely within the past few decades, and there is a high degree of confidence that it can be continued safely for limited periods of time. However, as the amounts of radioactive waste in surface storage have increased, concern has grown over the sustainability of storage in the long term and the associated safety and security implications. In response to these concerns, the IAEA has prepared a position paper [1] that is intended for general readership. This presentation will provide a summary of the position paper, and a discussion of some safety issues for further consideration. A key theme is the contrast of the safety and sustainability implications of long term storage with those of early disposal. A number of factors are examined from different points of view, factors such as safety and security, need of maintenance, institutional control and information transfer, community attitudes and availability of funding. The timing and duration of the process of moving from storage to disposal, which are influenced by factors such as the long timeframes required to implement disposal and changing public attitudes, will also be discussed. The position paper focuses on the storage of three main types of waste: high level waste from the reprocessing of nuclear fuel, spent nuclear fuel that is regarded as waste and long-lived intermediate level radioactive waste. Long term storage of mining and milling waste, and other large volumes of waste from processes involving the use of naturally occurring radioactive materials are not discussed. Specialist meetings were held last year by the IAEA on the sustainability and safety of long-term storage to establish and discuss the issues where a broad consensus exists, and to investigate areas where issues remain unresolved. Within the technical community, it is widely agreed that perpetual storage is not considered to be either feasible or acceptable because of the impossibility of assuring active control over the time periods for which these wastes remain potentially hazardous. For high-level and long-lived radioactive waste, the consensus of the waste management experts is that disposal in deep underground engineered facilities — geological disposal — is the best option that is currently available, or likely to be available in the foreseeable future.

Significance of long-term climate evolution and associated impacts on the long-term safety of a high-level radioactive waste repository within the German siting process

2021 ◽  
Author(s):  
Marc Wengler ◽  
Astrid Göbel ◽  
Eva-Maria Hoyer ◽  
Axel Liebscher ◽  
Sönke Reiche ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
Site Selection ◽  
Phase 1 ◽  
High Level Radioactive Waste ◽  
Safety Requirements ◽  
Climate Evolution ◽  
Safety Assessments ◽  
High Level ◽  
Host Rocks

<p>According to the 'Act on the Organizational Restructuring in the Field of Radioactive Waste Disposal' the BGE was established in 2016. The amended 'Repository Site Selection Act' (StandAG) came into force in July 2017 and forms the base for the site selection by clearly defining the procedure. According to the StandAG the BGE implements the participative, science-based, transparent, self-questioning and learning procedure with the overarching aim to identify the site for a high-level radioactive waste (HLW) repository in a deep geological formation with best possible safety conditions for a period of one million years.</p><p>The German site selection procedure consists of three phases, of which Phase 1 is divided into two steps. Starting with a blanc map of Germany, the BGE completed Step 1 in September 2020 and identified 90 individual sub-areas that provide favorable geological conditions for the safe disposal of HLW in the legally considered host rocks; rock salt, clay and crystalline rock. Based on the results of Step 1, the on-going Step 2 will narrow down these sub-areas to siting regions for surface exploration within Phase 2 (§ 14 StandAG). Central to the siting process are representative (Phase 1), evolved (Phase 2) and comprehensive (Phase 3) preliminary safety assessments according to § 27 StandAG.</p><p>The ordinances on 'Safety Requirements' and 'Preliminary Safety Assessments' for the disposal of high-level radioactive waste from October 2020 regulate the implementation of the preliminary safety assessments within the different phases of the siting process. Section 2 of the 'Safety Requirements' ordinance provides requirements to evaluate the long-term safety of the repository system; amongst others, it states that all potential effects that may affect the long-term safety of the repository system need to be systematically identified, described and evaluated as “expected” or “divergent” evolutions. Additionally, the ordinance on 'Preliminary Safety Assessments' states in § 7, amongst others, that the geoscientific long-term prediction is a tool to identify and to evaluate geogenic processes and to infer “expected” and “divergent” evolutions from those. Hence, considering the time period of one million years for the safe disposal of the HLW and the legal requirements, it is essential to include long-term climate evolution in the German site selection process to evaluate the impact of various climate-related scenarios on the safety of the whole repository system.</p><p>To better understand and evaluate the influence of climate-related processes on the long-term safety of a HLW repository, climate-related research will be a part of the BGE research agenda. Potential research needs may address i) processes occurring on glacial – interglacial timescales (e.g. the inception of the next glaciation, formation and depth of permafrost, glacial troughs, sub-glacial channels, sea-level rise, orbital forcing) and their future evolutions, ii) effects on the host rocks and the barrier system(s) as well as iii) the uncertainties related to these effects but also to general climate models and predictions.</p>

Methodology in Corrosion Testing of Container Materials for Radioactive Waste Disposal in a Geological Clay Repository

2001 ◽  
Author(s):  
Bruno Kursten ◽  
Frank Druyts ◽  
Pierre Van Iseghem
Keyword(s):  
Radioactive Waste ◽  
Corrosion Behaviour ◽  
Open Circuit ◽  
Geological Disposal ◽  
Boom Clay ◽  
Engineered Barriers ◽  
High Level ◽  
Aerobic And Anaerobic

Abstract The current worldwide trend for the final disposal of conditioned high-level, medium-level and long-lived alpha-bearing radioactive waste focuses on deep geological disposal. During the geological disposal, the isolation between the radioactive waste and the environment (biosphere) is realised by the multibarrier principle, which is based on the complementary nature of the various natural and engineered barriers. One of the main engineered barriers is the metallic container (overpack) that encloses the conditioned waste. In Belgium, the Boom Clay sediment is being studied as a potential host rock formation for the final disposal of conditioned high-level radioactive waste (HLW) and spent fuel. Since the mid 1980’s, SCK•CEN has developed an extensive research programme aimed at evaluating the suitability of a wide variety of metallic materials as candidate overpack material for the disposal of HLW. A multiple experimental approach is applied consisting of i) in situ corrosion experiments, ii) electrochemical experiments (cyclic potentiodynamic polarisation measurements and monitoring the evolution of ECORR as a function of time), and iii) immersion experiments. The in situ corrosion experiments were performed in the underground research facility, the High Activity Disposal Experimental Site, or HADES, located in the Boom clay layer at a depth of 225 metres below ground level. These experiments aimed at predicting the long-term corrosion behaviour of various candidate container materials. It was believed that this could be realised by investigating the medium-term interactions between the container materials and the host formation. These experiments resulted in a change of reasoning at the national authorities concerning the choice of over-pack material from the corrosion-allowance material carbon steel towards corrosion-resistant materials such as stainless steels. The main arguments being the severe pitting corrosion during the aerobic period and the large amount of hydrogen gas generated during the subsequent anaerobic period. The in situ corrosion experiments however, did not allow to unequivocally quantify the corrosion of the various investigated candidate overpack materials. The main shortcoming was that they did not allow to experimentally separate the aerobic and anaerobic phase. This resulted in the elaboration of a new laboratory programme. Electrochemical corrosion experiments were designed to investigate the effect of a wide variety of parameters on the localised corrosion behaviour of candidate overpack materials: temperature, SO42−, Cl−, S2O32−, oxygen content (aerobic - anaerobic),… Three characteristic potentials can be derived from the cyclic potentiodynamic polarisation (CPP) curves: i) the open circuit potential, OCP, ii) the critical potential for pit nucleation, ENP, and iii) the protection potential, EPP. Monitoring the open circuit potential as a function of time in clay slurries, representative for the underground environment, provides us with a more reliable value for the corrosion potential, ECORR, under disposal conditions. The long-term corrosion behaviour of the candidate overpack materials can be established by comparing the value of ECORR relative to ENP and EPP (determined from the CPP-curves). The immersion tests were developed to complement the in situ experiments. These experiments aimed at determining the corrosion rate and to identify the corrosion processes that can occur during the aerobic and anaerobic period of the geological disposal. Also, some experiments were elaborated to study the effect of graphite on the corrosion behaviour of the candidate overpack materials.

The Cige´o Industrial Geological Repository Project

2011 ◽  
Author(s):  
Thibaud Labalette ◽  
Alain Harman ◽  
Marie-Claude Dupuis
Keyword(s):  
Radioactive Waste ◽  
Phase 1 ◽  
Public Debate ◽  
Territorial Development ◽  
The Public ◽  
The Future ◽  
Geological Formation ◽  
Local Actors ◽  
High Level

The Planning Act of 28 June 2006 prescribed that a reversible repository in a deep geological formation be chosen as the reference solution for the long-term management of high-level and intermediate-level long-lived radioactive waste. It also entrusted the responsibility of further studies and investigations on the siting and design of the new repository upon the French Radioactive Waste Management Agency (Agence nationale pour la gestion des de´chets radioactifs – Andra), in order for the review of the creation-licence application to start in 2015 and, subject to its approval, the commissioning of the new repository in 2025. In late 2009, Andra submitted to the French government proposals concerning the implementation and the design of Cige´o (Centre industriel de stockage ge´ologique). A significant step of the project was completed with the delineation of an interest zone for the construction of the repositor’s underground facilities in 2010. This year, Andra has launched a new dialogue phase with local actors in order to clarify the implementation scenarios on the surface. The selected site will be validated after the public debate that is now scheduled for the first half of 2013. This debate will be organized by the National Public Debate Committee (Commission nationale du de´bat public). In parallel, the State is leading the preparation of an territorial development scheme, which will be presented during the public debate. The 2009 milestone also constitutes a new step in the progressive design process of the repository. After the 1998, 2001 and 2005 iterations, which focused mainly on the long-term safety of the repository, the Dossier 2009 highlighted its operational safety, with due account of the non-typical characteristics of an underground nuclear facility. It incorporates the first results of the repository-optimisation studies, which started in 2006 and will continue in the future. The reversibility options for the repository constitute proposals in terms of added flexibility in repository management and in package-recovery levels. They orient the design of the repository in order to promote those reversibility components. They contribute to the dialogue with stakeholders in the preparation of the public debate and of the future act on the reversibility conditions of the repository. The development of the repository shall be achieved over a long period, around the century. Hence, the designer will acquire additional knowledge at every new development of the project, notably during Phase 1, which he may reuse during the following phase, in order, for instance, to optimise the project. This process is part of the approach proposed by Andra in 2009 pursuant to the reversibility principle.

scholarly journals The Disposal of Radioactive Waste in Ice Sheets

1977 ◽  
Vol 19 (81) ◽  
pp. 607-617 ◽  
Author(s):  
K. Philberth
Keyword(s):  
Radioactive Waste ◽  
Ice Sheets ◽  
Ice Sheet ◽  
Climatic Changes ◽  
High Level Waste ◽  
Depth Range ◽  
The Stability ◽  
High Level ◽  
The Antarctic ◽  
Green Land

AbstractThe waste containers should be retrievable for a few centuries until further research has solved all problems and 90Sr and 137Cs have decayed to less than 0.1%. Safe and fairly cheap retrievability can be guaranteed without container mooring. The paper presents an example: The high-level waste of the whole world for the next 30 years could be put in to 3 × 107 spherical containers with 0.2 m radius and disposed of in an area with 15 km radius and a depth range of 20–100 m under the surface of either the Antarctic or the Green land ice sheet. The deposit does not affect the stability of the sheet. Even the most upsetting natural ice-sheet instabilities and/or climatic changes could not cause radioactive contamination.

scholarly journals Effect of Fe2O3 on the Immobilization of High-Level Waste with Magnesium Potassium Phosphate Ceramic

2019 ◽  
Vol 2019 ◽  
pp. 1-10
Author(s):  
Hailin Yang ◽  
Mingjiao Fu ◽  
Bobo Wu ◽  
Ying Zhang ◽  
Ruhua Ma ◽  
...  
Keyword(s):  
Sintering Temperature ◽  
Potassium Phosphate ◽  
High Level Waste ◽  
Radioactive Nuclide ◽  
Hydration Reaction ◽  
Sintering Process ◽  
Positive Role ◽  
Ceramic Structure ◽  
High Level

For the proposed novel procedure of immobilizing HLW with magnesium potassium phosphate cement (MKPC), Fe2O3 was added as a modifying agent to verify its effect on the solidification form and the immobilization of the radioactive nuclide. The results show that Fe2O3 is inert during the hydration reaction. It slows down the hydration reaction and lowers the heat release rate of the MKPC system, leading to a 3°C-5°C drop in the mixture temperature during hydration. Early comprehensive strength of Fe2O3 containing samples decreased slightly while the long-term strength remained unchanged. For the sintering process, Fe2O3 played a positive role, lowering the melting point and aiding the formation of ceramic structure. CsFe(PO4)2, or CsFePO4, was generated by sintering at 900°C. These products together with the ceramic structure and absorption benefit the immobilization of Cs+. The optimal sintering temperature for heat treatment is 900°C; it makes the solidification form a fired ceramic-like structure.

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