WAVELETS AND STATISTICAL TECHNIQUES FOR DATA ANALYSIS IN A MOCK-UP HIGH-LEVEL WASTE STORAGE EXPERIMENT

2004 ◽  
Vol 02 (04) ◽  
pp. 351-370 ◽  
Author(s):  
I. CAÑAMÓN ◽  
F. J. ELORZA ◽  
A. MANGIN ◽  
P. L. MARTÍN ◽  
R. RODRÍGUEZ
Keyword(s):  
Radioactive Waste ◽  
Large Scale ◽  
Matching Pursuit ◽  
High Level Waste ◽  
Data Series ◽  
Distribution Analysis ◽  
Physical Processes ◽  
Statistical Time ◽  
Correlation And Spectral Analysis ◽  
High Level

This work analyzes the physical processes occurring in the Mock-up test of the FEBEX I and II projects. FEBEX I and II is an European research project (1996–2004) led by ENRESA, that has financial support from the European Commission. This experiment is based in two large-scale heating tests ("in-situ" test and "Mock-up" test) simulating a radioactive waste repository, and tries to analyze the thermo-hydro-mechanical (THM) processes that could eventually happen in this kind of repositories.The main objectives of this study have been the following: to identify the physical processes occurring in the Mock-up experiment and to characterize them quantitatively; to understand the nature and consequences of several incidents happening in the Mock-up during the heating phase; and, finally, to analyze the data reliability of the sensors measurements and to predict possible failures.The analysis techniques used in this work are both statistical (time series correlation and spectral analysis, wavelets analysis, matching pursuit analysis) and non-statistical (spatial distribution analysis of data). These methods aim to establish the existing relationships between several data series registered in the experiment, corresponding to the measured parameters, and to characterize on time and frequency the non-stationary response of the series. A better understanding of the main THM processes affecting the engineered barriers used for the isolation of the radioactive waste is then available with the results obtained from those analyses.

The Belgian Demonstration Programme for the Disposal of High-Level and Long-Lived Radioactive Waste

2006 ◽  
Vol 932 ◽  
Author(s):  
Bernier Frédéric ◽  
Demarche Marc ◽  
Bel Johan
Keyword(s):  
Radioactive Waste ◽  
Thermal Load ◽  
Large Scale ◽  
Mechanical Response ◽  
Boom Clay ◽  
Geological Formation ◽  
Important Input ◽  
Clay Layers ◽  
High Level ◽  
Large Scale Tests

ABSTRACTThe EIG EURIDICE is responsible for performing large-scale tests, technical demonstrations and experiments so as to assess the feasibility of a final disposal of vitrified radioactive waste in deep clay layers. This programme is part of the Belgian Research and Development programme managed by ONDRAF/NIRAS. The research infrastructure includes the Underground Research Facilities HADES (URF HADES) in the Boom Clay geological formation and surface facilities. The achievements of the demonstration programme are the demonstration of the construction of shafts and galleries at industrial scale, the characterisation of the hydro-mechanical response of the host rock, and the “OPHELIE mock-up” a large scale hydration test under thermal load of pre-fabricated bentonite blocks. The future works will consist mainly in the realisation of the “PRACLAY experiments” including a large scale heater test. The large scale heater test has to demonstrate that Boom Clay is suitable, in terms of performance of the disposal system, to undergo the thermal load induced by the vitrified waste. The combined effect of the excavation and the thermal load will be investigated. A long term (more than 10 years) large scale heater test would be representative of the most penalizing conditions that could be encountered in the real disposal. The results of this test will constitute an important input for the Safety and Feasibility Cases 1 (SFC-1, 2013) and 2 (SFC-2, 2020).

Manufacturing and Handling Techniques of the Monolithic Buffer Material for HLW Disposal

2001 ◽  
Author(s):  
Toshihiko Awano ◽  
Takeshi Kanno ◽  
Susumu Kawakami ◽  
Hiroyoshi Ueda ◽  
Takahiro Kimoto
Keyword(s):  
Radioactive Waste ◽  
Large Scale ◽  
Cold Isostatic Pressing ◽  
Large Block ◽  
Technical Feasibility ◽  
Geological Disposal ◽  
Buffer Material ◽  
Manufacturing Tests ◽  
High Level ◽  
Large Scale Tests

Abstract Small and large scale tests were performed to evaluate technical feasibility of the monolithic buffer material, defined as a large block of bentonite formed by the cold isostatic pressing, for geological disposal of high-level radioactive waste. Trial manufacturing tests up to approximately 70 [%]-scale of a Japanese disposal concept were carried out and emplacement tests were carried out by vacuum lifting and forklift-type methods for vertical and horizontal emplacement concepts, respectively. Based on the large engineering-scale tests, technical feasibility of the monolithic buffer was demonstrated.

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.

A Review of Corrosion Considerations of Container Materials Relevant to Underground Disposal of High-Level Radioactive Waste in Belgium

2006 ◽  
Vol 932 ◽  
Author(s):  
Bruno Kursten ◽  
Frank Druyts
Keyword(s):  
Radioactive Waste ◽  
Large Scale ◽  
General Corrosion ◽  
Spent Fuel ◽  
High Level Radioactive Waste ◽  
Boom Clay ◽  
Nuclear Site ◽  
In Situ Experiments ◽  
Underground Disposal ◽  
High Level

ABSTRACTThe underground formation that is currently being considered in Belgium for the permanent disposal of high-level radioactive waste and spent fuel is a 30-million-year-old argillaceous sediment (Boom Clay layer). This layer is located in the northeast of Belgium and extending under the Mol-Dessel nuclear site at a depth between 180 and 280 meter.Within the concept for geological disposal (multibarrier system), the metallic container is the primary engineered barrier. Its main goal is to contain the radioactive waste and to prevent the groundwater from coming into contact with the wasteform by acting as a tight barrier. The corrosion resistance of container materials is an important aspect in ensuring the tightness of the metallic container and therefore plays an important role in the safe disposal of HLW. The metallic container has to provide a high integrity, i.e. no through-the-wall corrosion should occur, at least for the duration of the thermal phase (500 years for vitrified HLW and 2000 years for spent fuel).An extensive corrosion evaluation programme, sponsored by the national authorities and the European Commission, was started in Belgium in the mid 1980's. The main objective was to evaluate the long-term corrosion performance of a broad range of candidate container materials. In addition, the influence of several parameters, such as temperature, oxygen content, groundwater composition (chloride, sulphate and thiosulphate), γ-radiation, … were investigated. The experimental approach consisted of in situ experiments (performed in the underground research facility, HADES), electrochemical experiments, immersion experiments and large scale demonstration tests (OPHELIE, PRACLAY). Degradation modes considered included general corrosion, localised corrosion (pitting) and stress corrosion cracking.This paper gives an overview of the more relevant experimental results, gathered over the past 25 years, of the Belgian programme in the field of container corrosion.

Impact of Advanced Fuel Cycles on Geological Disposal

2009 ◽  
Vol 1193 ◽  
Author(s):  
Jan Marivoet ◽  
Eef Weetjens
Keyword(s):  
Radioactive Waste ◽  
Cooling Time ◽  
High Level Waste ◽  
Waste Streams ◽  
Geological Disposal ◽  
Fuel Cycles ◽  
Thermal Output ◽  
Advanced Fuel ◽  
High Level ◽  
The Impact

AbstractIn recent years the increasing oil prices and the need for carbon-free energy to limit global warming have resulted in a revival of interests in nuclear energy. Advanced nuclear fuel cycles are being studied worldwide. They aim at making more efficient use of the available resources, reducing the risk of proliferation of nuclear weapons, and facilitating the management of the resulting radioactive waste. Recently, the Red-Impact project has investigated the impact of a number of representative advanced fuel cycles on radioactive waste management, and more specific on geological disposal. The thermal output of the high-level waste arising from advanced fuel cycles in which all the actinides are recycled is reduced with a factor 3 for a 50 years cooling time and with a factor 5 for a 100 years cooling time in comparison with the spent fuel arising from the once-through fuel cycle. This reduction of the thermal output allows for a significant reduction of the length of the disposal galleries and of the size of the repository. Separation of Cs and Sr drastically reduces further the thermal output of the high-level waste, but it requires a long-term management of those heat generating separated waste streams, which contain the very long-lived 135Cs. Recycling all the actinides strongly reduces the radiotoxicity in the waste, resulting in significantly lower doses to an intruder in the case of a human intrusion into the repository. However, the reduction of radiotoxicity has little impact on the main safety indicator of a geological repository, i.e. the effective dose in the case of the expected evolution scenario; for disposal in clay formations, this dose is essentially due to mobile fission and activation products. The deployment of advanced fuel cycles will necessitate the development of low activation materials for the new nuclear facilities and fuels and of specific waste matrices to condition the high-level and medium-level waste streams that will arise from the advanced reprocessing plants.

Dry Storage of Spent Research Reactor Fuel in Castor BR3® Casks at Belgoprocess in Belgium

2001 ◽  
Author(s):  
Marnix Braeckeveldt ◽  
Luc Ooms ◽  
Gustaaf Geenen
Keyword(s):  
Radioactive Waste ◽  
Nuclear Research ◽  
High Level Waste ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Dry Storage ◽  
Nuclear Site ◽  
Test Reactor ◽  
Intermediate Storage ◽  
High Level

Abstract The BR3 reactor (10.5 MWe) at the Nuclear Research Center SCK•CEN was the first PWR plant installed in Europe and has been shut down in 1987. The BR3 reactor is from 1989 in a decommissioning stage and most of the spent fuel is presently still stored in the deactivation pool of the BR3 plant and has to be evacuated. The BR3 was used as a test-reactor for new fuel types and assemblies (Mixed Oxide (MOX) fuel, fuel rods containing burnable poison (Gd2O3) and other types of fuels). Some fuel rods, having undergone a destructive analysis, are stored in different laboratories at the SCK•CEN. In total, the BR3 spent fuel comprises the equivalent of almost 200 fuel assemblies corresponding to some 5000 fuel rods. Beside the spent BR3 fuel, a limited number of spent fuel rods, with equivalent characteristics as the BR3 fuel but irradiated in research reactors outside Belgium and stored in other buildings at the SCK•CEN nuclear site, were added to the inventory of spent fuel to be evacuated. Various options such as reprocessing and intermediate storage awaiting final disposal were evaluated against criteria as available techniques, safety, waste production and overall costs. Finally the option of an AFR (away-from-reactor) intermediate dry storage of the BR3 and other spent fuel in seven CASTOR BR3® casks was adopted. As the SCK•CEN declared this spent fuel as radioactive waste, NIRAS/ONDRAF, the Belgian radioactive waste management agency became directly involved and the decision was taken to construct a small building at the Belgoprocess nuclear site for storing the CASTOR BR3® casks. Loading at the SCK•CEN followed by transport to Belgoprocess and storage is scheduled to take place at the end of 2001. The CASTOR BR3® cask weighing some 25 tonnes, consists of a monolithic body and has two independent lids with metal seals guaranteeing the long term leak-tightness of the cask. The CASTOR BR3® cask is designed for transport and the intermediate storage of at least 50 years. Although a defect of the leaktightness of a CASTOR BR3® cask is very unlikely to occur, an intervention scenario had to be developed. As no pool is present at the Belgoprocess nuclear site to unload the fuel, an innovative procedure is developed that consists of transferring the basket, containing the spent fuel, into another CASTOR BR3® cask. This operation can be performed in the hot cell of the existing storage building for high level waste at the Belgoprocess site.

Shared, Regional Repositories: Developing a Practical Implementation Strategy

2009 ◽  
Author(s):  
Ewoud Verhoef ◽  
Charles McCombie ◽  
Neil Chapman
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Basic Concept ◽  
Implementation Strategy ◽  
Spent Nuclear Fuel ◽  
Action Plan ◽  
High Level Waste ◽  
Practical Implementation ◽  
Strategic Action ◽  
High Level

The basic concept within both EC funded SAPIERR I and SAPIERR II projects (FP6) is that of one or more geological repositories developed in collaboration by two or more European countries to accept spent nuclear fuel, vitrified high-level waste and other long-lived radioactive waste from those partner countries. The SAPIERR II project (Strategic Action Plan for Implementation of Regional European Repositories) examines in detail issues that directly influence the practicability and acceptability of such facilities. This paper describes the work in the SAPIERR II project (2006–2008) on the development of a possible practical implementation strategy for shared, regional repositories in Europe and lays out the first steps in implementing that strategy.

The Implementing Geological Disposal Technology Platform: Key Challenges in Research and Development in Radioactive Waste Management

2014 ◽  
Author(s):  
Jacques Delay ◽  
Jiri Slovak ◽  
Raymond Kowe
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Added Value ◽  
High Level Waste ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Technology Platform ◽  
Joint Activities ◽  
The Individual ◽  
High Level

The Implementing Geological Disposal of Radioactive Waste Technology Platform (IGD-TP) was launched in November 2009 to tackle the remaining research, development and demonstration (RD&D) challenges with a view to fostering the implementation of geological disposal programmes for high-level and long-lived waste in Europe. The IGD-TP’s Vision is that “by 2025, the first geological disposal facilities for spent fuel, high-level waste and other long-lived radioactive waste will be operating safely in Europe”. Aside from most of European waste management organisations, the IGD-TP now has 110 members covering most of the RD&D actors in the field of implementing geological disposal of radioactive waste in Europe. The IGD-TP Strategic Research Agenda (SRA), that defines shared RD&D priorities with an important cooperative added value, is used as a basis for the Euratom programme. It provides a vehicle to emphasise RD&D and networking activities that are important for establishing safety cases and fostering disposal implementation. As the IGD-TP brings together the national organisations which have a mandate to implement geological disposal and act as science providers, its SRA also ensures a balance between fundamental science, implementation-driven RD&D and technological demonstration. The SRA is in turn supported by a Deployment Plan (DP) for the Joint Activities to be carried out by the Technology Platform with its members and participants. The Joint Activities were derived from the individual SRA Topics and prioritized and assigned a timeline for their implementation. The deployment scheme of the activities is updated on a yearly basis.

Creation of system of final isolation (disposal) of radioactive waste in the Russian Federation

MRS Advances ◽  
2020 ◽  
Vol 5 (5-6) ◽  
pp. 275-282 ◽  
Author(s):  
Vsevolod Igin ◽  
Victor Krasilnikov
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Waste Disposal ◽  
Russian Federation ◽  
Radioactive Waste Disposal ◽  
High Level Waste ◽  
Federal State ◽  
Radioactive Waste Management ◽  
Near Surface ◽  
High Level

Abstract:The paper provides generic overview of legal and regulatory framework of radioactive waste management activities held in Russian Federation and national operator responsibilities and accomplishments. It gives a short description of waste classification scheme used and plans for radioactive waste disposal. In particular the paper provides information on the plans of the FEDERAL STATE UNITARY ENTERPRISE "National operator for radioactive waste management" to construct and operate several near-surface disposal facilities for low and intermediate level waste with total capacity up to 550 000 cubic meter. The paper also provides detailed information on the steps of high-level waste disposal program including site-selection, construction phase of the underground research laboratory (URL) near the city of Zheleznogorsk, Krasnoyarsk Region and research program after the construction of the URL. The paper also describes Russian system and state policy in the field of RW management and gives recommendations for future implementers.

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.

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