scholarly journals Deep borehole disposal of intermediate-level waste

2021 ◽  
Vol 1 ◽  
pp. 263-264
Author(s):  
Dirk Mallants ◽  
John Phalen ◽  
Hef Griffiths
Keyword(s):  
Spent Nuclear Fuel ◽  
Cost Effective ◽  
Full Scale ◽  
Intermediate Level ◽  
High Level Waste ◽  
Deep Borehole ◽  
International Partnership ◽  
Geologic Disposal ◽  
Safety Assessments

Abstract. Around the world, deep borehole disposal is being evaluated for intermediate-level waste (ILW), high-level waste (HLW), spent nuclear fuel (SNF), separated plutonium waste and some very high specific activity fission product waste. In Australia, long-lived ILW from research reactors and radiopharmaceutical production represents the principal waste stream that requires deep geologic disposal. Whilst the Australian government has not yet made a decision on its preferred strategy for ILW disposal, deep borehole disposal of small volumes of ILW would be a more cost-effective and modular solution compared to a conventional geologic disposal facility (GDF). CSIRO, ANSTO and SANDIA have created an international partnership to execute a full-scale borehole research, development and demonstration (RD&D) project in Australia. The project will demonstrate the technical feasibility of the long-term safety of borehole disposal in deep geological formations. The execution of this project could also demonstrate options for nuclear waste disposal that would reduce proliferation risks, potentially up to the termination of compliance with international safeguards requirements. The RD&D includes demonstration of surface handling and waste/seal emplacement capabilities, basic research on foundational science areas, and full-scale field testing in both a deep characterization borehole and a larger-diameter (0.7 m or 27.5 inch) 2000 m deep demonstration borehole. The multi-barrier system designed for such a deep disposal borehole concept places much less reliance on engineered barriers at the disposal zone to achieve safety as compared to a conventional GDF. It rather relies on geological features for waste containment. The concept being explored uses disposal containers with primary waste packages, such as vitrified waste canisters, inside; to be both cost effective and fit for purpose, such a container could have a mild steel-based structural component with copper coating. A critical review of six coating technologies showed that cold spray has the greatest advantages, such as minimal porosity and compressive residual stress. The RD&D has delivered novel enabling tools that assist with site screening, borehole design and post-closure safety assessments. For instance, an automated geological fault mapping and meshing tool was developed that assists with ranking the suitability of potential disposal sites based on proximity to faults. New codes were developed for better representation of fault zones in 2D/3D numerical flow and transport models, while also being more efficient to execute. Post-closure safety assessments tested the sensitivity of long-term safety with respect to disposal depth, rock permeability and sorption. Heat transport calculations explored the sensitivity of temperature evolution within the borehole to parameters such as heat load, borehole depth, geothermal gradients and rock thermal conductivity. For verification of host rock tightness while also demonstrating the absence of recent groundwater, a new noble gas analytical facility has been established for measuring rare noble gases in mineral fluid inclusions as indicators of very old pore fluids.

Testing and Numerical Simulation of Elastomeric Seals Under Consideration of Time Dependent Effects

2016 ◽  
Author(s):  
Mike Weber ◽  
Anja Kömmling ◽  
Matthias Jaunich ◽  
Dietmar Wolff ◽  
Uwe Zencker ◽  
...  
Keyword(s):  
Finite Element ◽  
Spent Nuclear Fuel ◽  
Viscoelastic Behavior ◽  
Element Model ◽  
Intermediate Level ◽  
High Level Waste ◽  
Spent Fuel ◽  
Safety Issues ◽  
Elastomeric Seals

Due to delays in the siting procedure to establish a deep geological repository for spent nuclear fuel and high level waste and in construction of the already licensed Konrad repository for low and intermediate level waste, extended periods of interim storage will become more relevant in Germany. BAM is involved in most of the cask licensing procedures and is responsible for the evaluation of cask-related long-term safety issues. Elastomeric seals are widely used as barrier seals for containers for low and intermediate level radioactive waste. In addition they are also used as auxiliary seals in spent fuel storage and transportation casks (dual purpose casks (DPC)). To address the complex requirements resulting from the described applications, BAM has initiated several test programs for investigating the behavior of elastomeric seals. These include experiments concerning the hyperelastic and viscoelastic behavior at different temperatures and strain rates, the low temperature performance down to −40°C, the influence of gamma irradiation and the aging behavior. The first part of the paper gives an overview of these tests, their relevant results and their possible impact on BAM’s work as a consultant in the framework of approval and licensing procedures. The second part presents an approach of the development of a finite element model using the finite element code ABAQUS®. The long-term goal is to simulate the complex elastomeric behavior in a complete lid closure system under specific operation and accident conditions.

Characteristics of Cementitious Paste for use in Deep Borehole Disposal of Spent Fuel and High Level Wasteforms

2015 ◽  
Vol 1744 ◽  
pp. 205-210 ◽  
Author(s):  
Nick C Collier ◽  
Karl P Travis ◽  
Fergus G F Gibb ◽  
Neil B Milestone
Keyword(s):  
Spent Nuclear Fuel ◽  
High Level Waste ◽  
Hardened Cement ◽  
Oil Well ◽  
Deep Borehole ◽  
Sodium Gluconate ◽  
Spent Fuel ◽  
High Temperature And Pressure ◽  
Temperature And Pressure ◽  
High Level

ABSTRACTDeep borehole disposal (or DBD) is now seen as a viable alternative to the (comparatively shallow) geologically repository concept for disposal of high level waste and spent nuclear fuel. Based on existing oil and geothermal well technologies, we report details of investigations into cementitious grouts as sealing/support matrices (SSMs) for waste disposal scenarios in the DBD process where temperatures at the waste package surface do not exceed ∼190ºC. Grouts based on Class G oil well cements, partially replaced with silica flour, are being developed, and the use of retarding admixtures is being investigated experimentally. Sodium gluconate appears to provide sufficient retardation and setting characteristics to be considered for this application and also provides an increase in grout fluidity. The quantity of sodium gluconate required in the grout to ensure fluidity for 4 hours at 90, 120 and 140°C is 0.05, 0.25 and 0.25 % by weight of cement respectively. A phosphonate admixture only appears to provide desirable retardation properties at 90°C. The presence of either retarder does not affect the composition of the hardened cement paste over 14 days curing and the phases formed are durable under conditions of high temperature and pressure.

Radionuclide Inventory of Vitrified Waste After Spent Nuclear Fuel Reprocessing at La Hague

2013 ◽  
Author(s):  
A. Meleshyn ◽  
U. Noseck
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Auxiliary Information ◽  
Opalinus Clay ◽  
High Level Waste ◽  
Stable Elements ◽  
High Level ◽  
Model Approach

The primary aim of the present work was to determine the inventories of the radionuclides and stable elements in vitrified high-level waste produced at La Hague and delivered to Germany, which are of importance for long-term safety assessment of final repositories for radioactive wastes. For a subset of these radionuclides and stable elements, the inventories were determined — either by direct measurements or by involving established correlations — and reported by AREVA. This allowed verification of the validity of application of a model approach utilizing the data of burnup and activation calculations and auxiliary information on the reprocessing and vitrification process operated at La Hague. Having proved that such a model approach can be applied for prediction of inventories of actinides, fission and activation products in vitrified waste, the present work estimated the minimum, average and maximum inventories of the radionuclides, which are of importance for long-term safety assessment of final repositories for radioactive waste but were not reported by AREVA for delivered CSD-V canisters. The average and maximum inventories in individual CSD-V canisters predicted in the present approach were compared to the inventories predicted by Nagra for canisters with vitrified waste delivered from La Hague to Switzerland [1]. This comparison revealed a number of differences between these inventories despite the fact that the canisters delivered to Switzerland were produced in essentially the same way and from the common reprocessing waste stock as CSD-V canisters delivered to Germany. Therefore, a further work is required in order to identify the reason for the discrepancy in the present estimation versus the Nagra estimation [1]. Such a work should also address the recommendation by the international peer review of the Safety Report of the Project Opalinus Clay to obtain estimates of the inventories of long-lived mobile radionuclides (such as 14C, 36Cl, 79Se, and 129I), which contribute most to the dose estimates in the radiological safety assessments, if possible, in agreement with other countries with similar waste streams in order for a coordinated set of data to be generated [2]. Since vitrified waste from reprocessing of spent nuclear fuel at La Hague was delivered to several countries — Belgium, France, Germany, Japan, Netherlands, and Switzerland — an international effort can be recommended.

Thermal Analysis for Geologic Codisposal of SNF Packages

2002 ◽  
Author(s):  
Si Y. Lee
Keyword(s):  
Thermal Analysis ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Peripheral Region ◽  
Waste Form ◽  
Temperature History ◽  
High Level Waste ◽  
Geologic Repository ◽  
Geologic Disposal ◽  
High Level

The engineering viability of disposal of aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) in a geologic repository requires a thermal analysis to provide the temperature history of the waste form. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal system and as input to assess the chemical and physical behavior of the waste form within the Waste Package (WP). The leading codisposal WP design proposes that a central DOE Al-SNF canister be surrounded by five Defense Waste Process Facility (DWPF) glass log canisters, that is, High-level Waste Glass Logs (HWGL’s), and placed into a WP in a geologic disposal system. A DOE SNF canister having about 0.4318m diameter is placed along the central horizontal axis of the WP. The five HWGL’s will be located around the peripheral region of the DOE SNF canister within the cylindrical WP container. The codisposal WP will be laid down horizontally in a drift repository. In this situation, two waste form options for Al-SNF disposition are considered using the codisposal WP design configurations. They are the direct Al-SNF form and the melt-dilute ingot. In the present work, the reference geologic and design conditions are assumed for the analysis even though the detailed package design is continuously evolved. This paper primarily dealt with the thermal performance internal to the codisposal WP for the qualification study of the WP containing Al-SNF. Thermal analysis methodology and decay heat source terms have been developed to calculate peak temperatures and temperature profiles of Al-SNF package in the DOE spent nuclear fuel canister within the geologic codisposal WP.

The IAEA Coordinated Research Programme on the Performance of High-Level Waste Forms and Packages Under Repository Conditions

1992 ◽  
Vol 294 ◽  
Author(s):  
Vladimir S. Tsyplenkov
Keyword(s):  
Near Field ◽  
Research Programme ◽  
High Level Waste ◽  
Future Research ◽  
Spent Fuel ◽  
Waste Forms ◽  
Safety Assessments ◽  
High Level ◽  
Coordination Meeting

ABSTRACTThe IAEA initiated, in 1991, a Coordinated Research Programme (CRP), with the aim of promoting the exchange of information on the results obtained by different countries in the performance of high-level waste forms and waste packages under conditions relevant to final repository. These studies are being undertaken to obtain reliable data as input to safety assessments and environmental impact analyses, for final disposal purposes. The CRP includes studies on waste forms that are presently of interest worldwide: borosilicate glass, Synroc and spent fuel.Ten laboratories leading in investigation of high-level waste form performance have already joined the programme. The results of their studies and plans for future research were presented at the first Research Coordination Meeting, held in Karlsruhe, Germany, in November 1991. The technical contributions concentrated on effecting an understanding of dissolution mechanisms of waste forms under simulated repository conditions. A quantitative interpretation of the chemical processes in the near field is considered a prerequisite for long-term predictions and for the formulation of a "source term" for performance assessment studies.

Testing and Numerical Simulation of Elastomers: From Specimen Tests to Simulation of Seal Behavior Under Assembly Conditions

2017 ◽  
Author(s):  
Mike Weber ◽  
Anja Kömmling ◽  
Matthias Jaunich ◽  
Dietmar Wolff ◽  
Uwe Zencker ◽  
...  
Keyword(s):  
Finite Element ◽  
Spent Nuclear Fuel ◽  
Circular Cross Section ◽  
Computer Code ◽  
Intermediate Level ◽  
Sensitivity Analyses ◽  
Spent Fuel ◽  
Test Configuration ◽  
Waste Container

Extended periods of interim storage are more relevant in Germany due to delays in the siting procedure to establish a deep geological repository for spent nuclear fuel, high level radioactive waste and in low/intermediate level waste container storage designated for the Konrad repository. BAM is involved in national cask licensing procedures and responsible for the evaluation of cask-related long-term safety issues. The long-term performance of elastomer seals for lid systems of transport and storage casks, whether used as auxiliary seals in spent fuel casks or as primary seals for low and intermediate level waste packages, is an important issue in this context. The polymeric structure of these seals causes a complex mechanical behavior with time-dependent sealing force reduction. The results of a comprehensive purpose-designed test program consisting of basic compression and tension tests as well as relaxation tests on unaged specimens of representative types of elastomers (fluorocarbon rubber (FKM) and ethylene propylene diene rubber (EPDM)) at different temperatures and strain rates are presented. They were used to identify the constitutive behavior and to obtain parameters for finite element material models provided by the computer code ABAQUS®. After estimating the influence of parameters such as Poisson’s ratio and friction coefficient by sensitivity analyses, the chosen values for the finite element simulation were validated by comparison with specimen test results. Based on this preliminary work the simulation of a specific laboratory test configuration containing a typical elastomer seal with circular cross section is presented. The chosen finite element material model and the related parameters had to show that they are able to represent not only the specimen behavior under predominantly uniaxial load but also the more complex stress states in real components. Deviations between the measured and calculated results are pointed out and discussed. The results from this work will be utilized in future simulations of aged elastomer behavior.

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.

scholarly journals Deep Borehole Disposal Safety Case

Energies ◽  
2019 ◽  
Vol 12 (11) ◽  
pp. 2141 ◽  
Author(s):  
Geoff A. Freeze ◽  
Emily Stein ◽  
Patrick V. Brady ◽  
Carlos Lopez ◽  
David Sassani ◽  
...  
Keyword(s):  
Reactive Transport ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Deep Borehole ◽  
Safety Case ◽  
Key Issues ◽  
Safety Assessments ◽  
Engineered Barriers ◽  
Technical Basis ◽  
Safety Strategy

The safety case for deep borehole disposal of nuclear wastes contains a safety strategy, an assessment basis, and a safety assessment. The safety strategy includes strategies for management, siting and design, and assessment. The assessment basis considers site selection, pre-closure, and post-closure, which includes waste and engineered barriers, the geosphere/natural barriers, and the biosphere and surface environment. The safety assessment entails a pre-closure safety analysis, a post-closure performance assessment, and confidence enhancement analyses. This paper outlines the assessment basis and safety assessment aspects of a deep borehole disposal safety case. The safety case presented here is specific to deep borehole disposal of Cs and Sr capsules, but is generally applicable to other waste forms, such as spent nuclear fuel. The safety assessments for pre-closure and post-closure are briefly summarized from other sources; key issues for confidence enhancement are described in greater detail. These confidence enhancement analyses require building the technical basis for geologically old, reducing, highly saline brines at the depth of waste emplacement, and using reactive-transport codes to predict their movement in post-closure. The development and emplacement of borehole seals above the waste emplacement zone is also important to confidence enhancement.

Safety Assessment Input for Site Selection: The Swedish Example

2011 ◽  
Author(s):  
Johan Andersson
Keyword(s):  
Spent Nuclear Fuel ◽  
Mineral Resources ◽  
Thermally Induced ◽  
Environment And Health ◽  
Site Characteristics ◽  
Calculated Risk ◽  
Flow Evolution ◽  
Safety Assessments ◽  
Groundwater Composition

Svensk Ka¨rnbra¨nslehantering AB (SKB) has performed comprehensive investigations of two candidate sites for a final repository for Sweden’s spent nuclear fuel. In March 2011 SKB decided to submit licence applications for a final repository at Forsmark. Before selection, SKB stated that the site that offers the best prospects for achieving long-term safety in practice would be selected. Based on experiences previous safety assessments, a number of issues related to long-term safety need to be considered in the context of site comparison. The factors include sensitivity to climate change such as periods of permafrost and glaciations, rock mechanics evolution including the potential for thermally induced spalling and sensitivity to potential future earthquakes, current and future groundwater flow, evolution of groundwater composition and proximity to mineral resources. Each of these factors related to long-term safety for the two candidate sites is assessed in a comparative analysis of site characteristics. The assessment also considers differences in biosphere conditions and in the confidence of the site descriptions. The comparison is concluded by an assessment on how the identified differences would affect the estimated radiological risk from a repository located at either of the sites. The assessment concludes that there are a number of safety related site characteristics for which the analyses do not show any decisive differences in terms of implications on safety, between the sites Forsmark and Laxemar. However, the frequency of water conducting fractures at repository depth is much smaller at Forsmark than at Laxemar. This difference, in turn, affects the future stability of the current favourable groundwater composition, which combined with the much higher flows at Laxemar would, for the current repository design, lead to a breach in the safety functions for the buffer and the canister for many more deposition positions at Laxemar than at Forsmark. Thereby the calculated risk for Forsmark will be considerably lower than that for Laxemar. What decided the choice is that Forsmark is thus mainly that it was judged to offer better prospects for achieving long-term safety in the final repository. Other factors considered included implications regarding repository construction and operation, difference in the footprint of the repository, comparisons regarding environment and health as well as social resources and local support.

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