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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

A study of transfer of UNF from non-disposable canisters - 15388

A study was performed by AREVA and URS for the Department of Energy (DOE) [1] to establish optimal methods for the opening of dry storage canisters (DSC) and the subsequent repackaging of used nuclear fuel (UNF) from the DSCs into generic standard transportation, aging, and disposal canisters (STADs) in wet or dry environments at reactor, consolidated storage, or repository sites. Eight dry and wet cutting technologies were evaluated: (1) skiving; (2) milling; (3) arc gouging/plasma torch; (4) cutting wheel; (5) diamond cable cutting; (6) laser cutting; (7) sawing; and (8) electrical discharge machining. The evaluated UNF repackaging technologies options included: (1) wet transfer in a pool; (2) dry transfer in a dry transfer system (DTS); (3) dry transfer in a fixed hot cell (FHC); (4) dry transfer in a mobile hot cell (MHC), and (5) a hybrid approach utilizing wet transfer in a pool and dry transfer in a FHC. A systems engineering evaluation was utilized for ranking the eight viable cutting options and the five different repackaging approaches for each of the sites. The systems engineering evaluation took into account six criteria (i.e., life cycle costs, worker dose, operational efficiency, technical maturity, ease of licensing, and waste generation) and performed a qualitative pairwise ranking of their relative importance to one another. Then quantitative or qualitative factors associated with each criterion (e.g., cost in millions of dollars, relative ease of licensing, etc.) for each option at each location (reactor, consolidated storage facility [CSF], repository) were developed based on historical experience, existing design studies, current 'state-of-the art', and/or engineering judgment. Using this input, the systems engineering evaluation identified dry skiving as the preferred cutting technology, with the wheel cutting technology in either a dry or wet environment identified as a promising technology that, with some RD and D, could mature to a better option (depending also on the disposition of the used DSC). For a reactor and a CSF/repository site, the systems engineering evaluation ranked the repackaging option using the skiving dry cutting technique and wet transfer of UNF in a pool as the highest option. However for a CSF and a repository site, the team recommended the hybrid option using the skiving dry cutting technique and both dry and wet transfer of UNF in a pool and FHC as a potentially better option (providing specific benefits such as higher throughput and more flexibility in operation). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Spent Nuclear Fuel Storage and Radioactive Waste Disposal in the United States: A Law and Policy Analysis - 20302

While Congress plays political football, spent nuclear fuel continues to sit in de facto interim nuclear waste storage sites throughout the country. This ad hoc approach - if it can even be called an approach - to nuclear waste management is in no one's best interests, because of financial, safety, environmental, and national security concerns with the status quo. Senate leadership had asserted that an up-or-down vote on the Yucca Mountain Project would take place in September or October 2019, but that did not occur. Rather, a conference bill was agreed to, and it provided that, although the nuclear program should receive some funding, funding was not made available for centralized interim storage ('CIS') or the Yucca Mountain Project. The President's FY2021 Budget Request does not contain any proposed funding for the Yucca Mountain Project ('the Project'). This paper will discuss the implications of that process, both for the Project itself and for the U.S. nuclear waste program more generally. In order to provide context for the current situation, this paper will begin by chronicling the history of U.S. nuclear waste management, from the passage of the Nuclear Waste Policy Act of 1982 and to the present day. It will discuss the roles of the U.S. Department of Energy, the Nuclear Regulatory Commission, federal and state political delegations, and the nuclear energy industry. The paper will then identify key themes that can be expected to affect the future of nuclear waste management in the United States. For example, it will consider whether development of additional CIS facilities is a viable short- to medium-term solution. CIS facilities may relieve immediate pressure at decommissioned or soon-to-be-decommissioned nuclear power plants; but with that pressure partially alleviated, support for building a long-term repository may be reduced. It will also discuss opposition by the State of New Mexico and others to the proposed New Mexico CIS project, based not only on siting concerns but also whether it might become a de facto repository, and whether the State of Texas and others may also be opposed, for similar reasons, to the proposed CIS facility in Texas. Additionally, the paper will examine consent-based siting and its implications for Nevada and other states that are potential hosts for spent nuclear fuel storage and disposal facilities. Further, if the Project is not a politically viable solution, the paper will examine whether it is now time for serious consideration of an alternate location for a permanent repository, as well as consideration of other options entirely, such as deep geologic disposal. The paper will conclude by looking ahead to the 2020 elections. It will describe the positions of candidates (particularly Presidential candidates) on spent nuclear fuel storage and disposal options, without advocating for any particular candidate or political party. Perhaps in 2020, we will at least see a start to a solution to the seemingly intractable problem of nuclear waste management in the United States. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Salt International Collaborations FY23 Update

This report summarizes the international collaborations conducted by Sandia funded by the US Department of Energy Office (DOE) of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D and Salt International work packages. This report satisfies the level-three milestone M3SF-23SN010303062. Several stand-alone sections make up this summary report, each completed by the participants. The sections discuss granular salt reconsolidation (KOMPASS), engineered barriers (RANGERS), numerical model comparison (DECOVALEX) and an NEA Salt Club working group on the development of scenarios as part of the performance assessment development process. Finally, we summarize events related to the US/German Workshop on Repository Research, Design and Operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Impact of Zr Content on Radionuclide Release Rates in Metal Waste Forms

Reference alloy waste forms (RAW) underwent hybrid corrosion/immersion testing to parameterize the ANL analytical oxidative-dissolution model to enable the calculation of fractional release rates. The waste forms tested at SRNL were variations of the RAW-3 formulation that uses HT9 as the main alloy component, and are meant to enable evaluation of the impact of Zr:U ratio on the release rates of actinides and Tc-99. The test solutions were air saturated alkaline and acidic brines, representing potential repositories with those conditions. The testing approach consisted of 4 major steps; 1) bare surface corrosion measurements, 2) hybrid potentiostatic hold/exposure measurements, 3) measurement of radionuclide concentrations and relations to anodic current from potentiostatic holds, and 4) identification of corroding phases using SEM/EDS of electrodes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Experimental determination of solubilities of magnesium borates: Solubility constants of boracite [Mg 3 B 7 O 13 Cl(cr)] and aksaite [MgB 6 O 7 (OH) 6 ·2H 2 O(cr)]

In this work, solubility measurements regarding boracite [Mg 3 B 7 O 13 Cl(cr)] and aksaite [MgB 6 O 7 (OH) 6 ·2H 2 O(cr)] from the direction of supersaturation were conducted at 22.5 ± 0.5 °C. The equilibrium constant (log 10 K 0 ) for boracite in terms of the following reaction, Mg 3 B 7 O 13 Cl(cr) + 15H 2 O(l) ⇌ 3Mg 2+ + 7B(OH) 4 – + Cl – + 2H + is determined as -29.49 ± 0.39 (2σ) in this study. The equilibrium constant for aksaite according to the following reaction, MgB 6 O 7 (OH) 6 •2H 2 O(cr) + 9H 2 O(l) ⇌ Mg 2+ + 6B(OH) 4 – + 4H + is determined as -44.41 ± 0.41 (2σ) in this work. This work recommends a set of thermodynamic properties for aksaite at 25 °C and 1 bar as follows: ΔH$0\atop{f}$ =-6063.70 ± 4.85 kJ·mol -1 , ΔG =-5492.55 ± 2.32 kJ·mol -1 , and S 0 = 344.62 ± 1.85 J·mol -1 ·K -1 . Among them, ΔG$0\atop{f}$ is derived from the equilibrium constant for aksaite determined by this study; ΔH$0\atop{f}$ is from the literature, determined by calorimetry; and S 0 is computed in the present work from ΔG$0\atop{f}$ and ΔH$0\atop{f}$. This investigation also recommends a set of thermodynamic properties for boracite at 25 °C and 1 bar as follows: ΔH$0\atop{f}$ =-6575.02 ± 2.25 kJ·mol -1 , ΔG$0\atop{f}$ =-6178.35 ± 2.25 kJ·mol -1 , and S 0 = 253.6 ± 0.5 J·mol -1 ·K -1 . Among them, ΔG$0\atop{f}$ is derived from the equilibrium constant for boracite determined by this study; S 0 is from the literature, determined by calorimetry; and ΔH$0\atop{f}$ is computed in this work from ΔG$0\atop{f}$ and S 0 . The thermodynamic properties determined in this study can find applications in many fields. For instance, in the field of material science, boracite has many useful properties including ferroelectric and ferroelastic properties. The equilibrium constant of boracite determined in this work will provide guidance for economic synthesis of boracite in an aqueous medium. Similarly, in the field of nuclear waste management, iodide boracite [Mg 3 B 7 O 13 I(cr)] is proposed as a waste form for radioactive 129 I. Therefore, the solubility constant for chloride boracite [Mg 3 B 7 O 13 Cl(cr)] will provide the guidance for the performance of iodide boracite in geological repositories. Boracite/aksaite themselves in geological repositories in salt formations may be solubility-controlling phase(s) for borate. Finally, solubility constants of boracite and aksaite will enable researchers to predict borate concentrations in equilibrium with boracite/aksaite in salt formations.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Sorption of Cationic Radionuclides on Chlorite and Sericite formed by the Hydrothermal Alteration - 18180

Chlorite and sericite formed by the hydrothermal alteration of granite distribute predominantly around the geological repository in deep underground. These minerals often cover the surface of a solid phase and fill the fracture of granite considered as the natural barrier to the migration of radionuclides leaching from radioactive waste. Therefore, to assess the practical performance of the natural barrier, it is necessary to clarify the interaction between radionuclides and these minerals. This study examined the sorption of cationic radionuclides by chlorite and sericite through batch experiments, whereby Cs{sup +}, Sr{sup 2+}, and Eu{sup 3+} were used as tracer ions. In these experiments, powder-form chlorite or sericite was mixed with the solutions of the tracer ions to a liquid/solid ratio of 10 ml/g, and the resulting mixture was allowed to incubate for 2-7 days. The concentration of tracer ions was set to 1 mM and that of NaCl to a specific value in the 0-0.6 M range. The solution pH was 3, 5, or 8. As a result, tracer ions were significantly adsorbed by chlorite and sericite, regardless of pH or NaCl concentration, except when NaCl concentration reached the high value of 0.6 M. The distribution coefficient for europium ions at pH 5 was estimated to be 2-7 mL/g in chlorite and 10{sup 3}-10{sup 4} mL/g in sericite. These values significantly exceeded that measured for feldspar (2-4 mL/g), the predominant component of granite before it undergoes any hydrothermal alteration. These results suggest that cationic radionuclides will be immobilized through sorption in chlorite and sericite filling granite fractures, even if these thermal-fluid-formed minerals cover the surface of granite. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

A Method for Understanding the Behaviour of Radioactive Decay Chains in Radiological Assessments - 19382

In many radiological groundwater pathway assessments, especially those where the assessment timescales of interest are relatively short (say a few thousand years or less), contributions to effective dose or radiological risk are dominated by radionuclides such as C-14, Cl-36 or I-129. Over longer timescales, additional radionuclides can also become key contributors to risk. In particular, members of the U-238 decay chain, notably Ra-226 and Pb-210, are likely to be key radionuclides. Other decay chains that can contribute significantly to dose or risk at times far into the future include Np-237 → U-233 → Th-229 and U-235 → Pa-231 → Ac-227. In this paper, a method is described for understanding decay chain behaviour in radiological assessments. The method uses a graphical approach in which the various modes of transport and decay in a system of connected compartments are represented as pathways between two nodes in a two-dimensional grid. Each pathway corresponds to a distinct transport and decay process. By so doing, the different transport and decay processes can be identified in a systematic way. The result is an algorithm for generating the contributions to effective dose or radiological risk without explicit solution of the transport equations. The method has been tested against numerical models and analytic solutions and found to give excellent agreement, while at the same time providing information about decay chain behaviour that would be difficult to obtain from a numerical assessment model. The power of the method is illustrated by applying it to some examples based on real repository concepts. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

The fourth decade of underground research at the Grimsel Test Site - what we have learned and where we go from here - 15075

At the Grimsel Test Site, NAGRA and its partners have celebrated thirty years' worth of essential contributions to the characterization of crystalline host rock, and the development and confirmation of safe geological disposal concepts. In operation since 1984, the underground research laboratory in the Swiss Alps shall not accept radioactive waste but can host experiments with radiotracers/radionuclides under strictly controlled conditions. The strong drive for international cooperation at the Grimsel Test Site is reflected in the more than two dozen organizations and research institutes from twelve countries and the European Union which have participated in the research program to date. The program has been designed dynamically to always mirror the current requirements of national programs for the long-term management of radioactive waste, as well as attempt to anticipate the next steps. Early work featured the development of methodologies for formation characterization. Then followed a period concentrating on performance assessment issues, before radionuclide migration and barriers were investigated intensely. The past 15 years have been dominated by engineering demonstration experiments. The research conducted at the Grimsel Test Site has complemented laboratory experiments, surface investigations and numerical modelling studies through the bridging of different scales and drawing off links between laboratory setups and natural analogues. The crystalline rock formations of the Aar Massif have provided an opportunity to do experiments in environments similar to those of future repositories. The facility also serves as a platform for continual knowledge transfer between generations of scientists and fosters synergies among disciplines. However, considering the ambitious schedule for repository implementation in several continuous, one may ask whether the era of generic underground rock laboratories is approaching its end. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Salt International Collaborations FY24 Update

This report summarizes the international collaborations conducted by Sandia funded by the US Department of Energy Office (DOE) of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D and Salt International work packages. This report satisfies the level-three milestone M3SF-24SN010303063. Several stand-alone sections make up this summary report, each completed by the participants. The sections discuss granular salt reconsolidation (KOMPASS/MEASURES), engineered barriers (RANGERS), numerical model comparison (DECOVALEX), an NEA Salt Club working group on the development of scenarios as part of the performance assessment development process, and progress on seal percolation experiments with German colleagues at Gesellschaft für Anlagen- und Reaktorsicherheit (GRS). Finally, we summarize events related to the US/German Workshop on Repository Research, Design and Operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Impact of lump characteristics on self-shielding corrections of active neutron interrogation on fissile material in waste - 15405

Non-destructive active neutron assay techniques such as active neutron coincidence counting, differential die away analysis or the californium shuffler method rely on the interrogation of fissile nuclear materials by an external neutron source. Whether it is a spontaneous fission or (α,n) neutron source, like {sup 252}Cf or AmLi; or a fusion neutron generator, all of these techniques rely on induced secondary fission events to provide prompt or delayed neutron signals that serve as the basis for the fissile material quantification. The accuracy of these fissile mass measurements is therefore dependent on the neutron interrogation efficiencies determined for calibration reference standards and whether the measured items are comparable to the calibration condition. In such favorable conditions, it is possible to establish a reliable empirical non-linear relationship between apparent (uncorrected) and real or true fissile mass. However, in waste measurements, this is unlikely to be the case; thus a calibration approach that takes account of the neutron self-shielding effect is required. By incorporating the Self-Shielding Factor (SSF) of the calibration reference standards, a linear mass calibration factor [s{sup -1}. {sup 239}Pu{sub eq} g{sup -1}] can be found; this calibration is accurate for waste in which the fissile material is 'infinitely dilute' i.e. the waste exhibits no self-shielding; this is assumed to be representative of dispersed material. The self-shielding correction and its systematic uncertainty are then left to be treated as part of the Total Measurement Uncertainty (TMU) evaluation associated with the assay result. The derivation of the TMU contribution associated with the self-shielding effect in nuclear waste is a complex problem for which a reliable evaluation has to be performed. This is usually based on assuming 'worst case scenarios' for the distribution of material that can arise in a particular waste stream. Such an approach is conservative, but often very penalizing, especially if one considers larger lumps (those containing grams of fissile material); for such lumps, the self-shielding correction becomes very significant, potentially altering the mass of {sup 239}Pu{sub eq} by over an order of magnitude. A worst case 'nuclear safety value' may not be the most appropriate choice for building a consignment best estimate or repository sentencing value for an item. In the present work the authors describe a numerical approach to study the variability of neutron self-shielding factors associated with distributions of a wide range of lump characteristics in terms of size, shape, chemical composition, density and isotopic composition. The described novel calculation tool is capable of accurately predicting the real impact of random distribution of lumps in a waste material, and allowed to show that given reasonable, defensible, assumptions on the composition of the lumps in waste stream, the systematic error component of the DDA TMU can be significantly reduced. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Criticality Consequence Analysis Roadmap for Spent Nuclear Fuel Canister in a Repository

This report discusses an initial criticality consequence analysis approach of commercial spent fuel canisters in a repository that will be developed to support a performance assessment (PA) of a repository with criticality events. A repository PA with criticality events may be needed to support direct disposal of currently loaded dual-purpose canisters. The ultimate goal is to develop a criticality consequence analysis model that can interface with any repository PA model. The criticality consequence analysis approach describes multi-physics coupling between neutronics, thermal hydraulics, and mechanics analysis codes. This report also discusses the sensitivity and uncertainty analysis approach for quantifying conservatism in the criticality consequence outcomes, as well as a defensible multi-physics validation approach to support any future licensing effort. This criticality consequence analysis approach will be refined in the future to support specific regulatory frameworks for a specific repository PA if needed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Report priority gaps in high temperature thermodynamic data (Interim Progress Report)

This interim progress report (Level 4 Milestone Number M4SF-26LL010203023) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Host Rock Properties & Processes SF-26LL01020302. Our focus is to assess gaps in data availability and understanding for radionuclide thermodynamics within the context of a “hot repository” concept and expand SUPCRT-NE database development to address higher temperatures needed for a DPC DGR disposal concept. The database is intended to inform the argillite GDSA baseline model. The leading European thermochemical database (Thermochimie) is only applicable to temperatures below 80°C. Thus, a US effort to integrate and expand upon other international thermodynamics database efforts is needed, particular if a “hot repository” concept moves forward.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Real-Time Electrochemical Measurements Of Carbon Steel In Ground Water With Sulfate Reducing Bacteria Enrichment - 15604

A new laboratory system for studying the microbiological corrosion of decommissioning waste was designed and developed during this project. The material in this part of the laboratory studies was carbon steel (AISI/SAE 1005/UNS G10050). In order to estimate the corrosive effects of microbial activity both biotic and abiotic (containing the biocides methyl-isothiazolinone and glutaraldehyde) mesocosms were prepared. To enhance the effects of sulfate reducing bacteria an enrichment of these bacteria isolated from the native drill hole water from the repository site was inoculated in to the biotic mesocosms. Microbiological corrosion was studied using new electrochemical technologies, such as multi-electrode arrays sensors, and molecular biology methods in the laboratory. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Safety assessment for the radon-type surface disposal facility at Saakadze, Georgia - 15406

The Saakadze disposal facility is located about 40 km from Tbilisi and was planned and constructed during Soviet times as Radon-type surface facility to accommodate up to 600 m{sup 3} of low and intermediate level waste generated on the territory of Georgia. The disposal facility was operated from 1963 until its closure in 1995. The vast majority of waste in the Saakadze facility has been disposed of in two near surface concrete vaults with typical Radon-type dimensions, which cover layers already show significant structural damages. In order to assess the present safety of the site and provide the competent Georgian authorities with the necessary basis to take decisions in regard to the future of the repository, the European Commission launched and financially supported a project, which was assigned to a consortium of TUV-Nord ENSYS GmbH and DBE Technology GmbH. The safety assessment of Saakadze facility, which was an essential part of this project and which is the subject of this paper was carried out by DBE Technology GmbH. The Government of Georgia has been gradually increasing efforts to improve the regulatory processes and the legislation on nuclear and radiation safety, and, in particular, on safe management of radioactive waste and sealed radioactive sources. Currently, however, there are still areas where specific Georgian laws and regulations do not exist. Accordingly, in certain aspects IAEA recommendations and best practices were used as guidelines for the methodology of the safety assessment. In spite of the large uncertainties to be considered, especially in regard to the inventory, preliminary results of the safety assessment indicate that scenarios related to the groundwater pathway do not lead to radiological impacts that are above or near to limiting values. However, human intrusion scenarios might lead to dose rates for the critical group that are exceeding these values. This is mainly due to the fact that today there is no proper engineered cover on top of the concrete vaults. Based on the comparison of existing options and considering the results of the long-term safety assessment, it is recommended to backfill the remaining voids inside the concrete vaults and to construct an engineered cover to reduce the probability of human intrusion scenarios and to minimize their potential impact. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Parameter Uncertainty for Repository Thermal Analysis

This report is one follow-on to a study of reference geologic disposal design concepts (Hardin et al. 2011a). Based on an analysis of maximum temperatures, that study concluded that certain disposal concepts would require extended decay storage prior to emplacement, or the use of small waste packages, or both. The study used nominal values for thermal properties of host geologic media and engineered materials, demonstrating the need for uncertainty analysis to support the conclusions. This report is a first step that identifies the input parameters of the maximum temperature calculation, surveys published data on measured values, uses an analytical approach to determine which parameters are most important, and performs an example sensitivity analysis. Using results from this first step, temperature calculations planned for FY12 can focus on only the important parameters, and can use the uncertainty ranges reported here. The survey of published information on thermal properties of geologic media and engineered materials, is intended to be sufficient for use in generic calculations to evaluate the feasibility of reference disposal concepts. A full compendium of literature data is beyond the scope of this report. The term “uncertainty” is used here to represent both measurement uncertainty and spatial variability, or variability across host geologic units. For the most important parameters (e.g., buffer thermal conductivity) the extent of literature data surveyed samples these different forms of uncertainty and variability. Finally, this report is intended to be one chapter or section of a larger FY12 deliverable summarizing all the work on design concepts and thermal load management for geologic disposal (M3FT-12SN0804032, due 15Aug2012).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

ENRESA's Spent Nuclear Fuel and Special Waste Database - 18115

Empresa Nacional de Residuos Radiactivos S.A. (ENRESA, by its initials in Spanish) Spent Nuclear Fuel (SNF) and Special Waste (SW) database development was initiated in a common effort between ENRESA and UNESA (the Spanish Association of the Electricity Industry) as a consequence of new regulatory requirements in the frame of radioactive waste management plans. Although these requirements were only mandatory for operators of nuclear power plants, ENRESA deemed necessary to have a centralized database which collected the data from the plant operators. The most relevant features of this database are: - The information collected covers not only the general characteristics of the SNF and SW but also the operation details needed for their characterization. - An application, named Gestion de Elementos Combustibles y Residuos Especiales (GECYRE, by its initials in Spanish) has been implemented for easing the queries and the generation of predefined reports. - An agile data transfer scheme, based on XML files, is used for an effective communication between ENRESA and the operators. The database is currently in production phase since 2014. During this phase it has demonstrated its effectiveness as an information repository to address the SNF and SW interim storage projects in Spain foreseen in the near future. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Overview of Canada's Plan for the Long-Term Management of Used Nuclear Fuel - 19537

In June 2007, the Government of Canada approved the Nuclear Waste Management Organisation's (NWMO) recommended Adaptive Phased Management (APM) approach for the long-term management of Canada's used nuclear fuel. As per this approach, used nuclear fuel will be isolated and contained in a centralized Deep Geological Repository (DGR) located in a stable crystalline or sedimentary rock formation at a nominal depth of 500 metres below ground surface. The DGR concept contains multiple barriers to safely contain and isolate the used nuclear fuel over the long-term. The used nuclear fuel bundles will be encapsulated in a copper-coated steel Used Fuel Container (UFC) and pre-packaged in a bentonite clay buffer box above ground. The UFC and bentonite clay buffer box assemblies will then be emplaced in excavated tunnel openings and sealed with clay-based backfill material and room seals, which completes the primary Engineered Barrier System (EBS). Since 2011, the NWMO has executed a design-development program to validate the design and manufacturing technologies for the EBS. Currently, the NWMO are executing a proof test program for the EBS for the purpose of demonstrating serial production. As the prototypes are produced, they will be subject to full-scale in-room emplacement trials within a custom-built above-ground mock emplacement room. The proof test program will be executed using APM reference designs, materials, processes, and procedures. An overview of this program is discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗