Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Repository”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 919 records · Page 51
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Investigation of Coupled Processes and Impact of High Temperature Limits in Argillite Rock: FY17 Progress. Predecisional Draft

The focus of research within the Spent Fuel and Waste Science and Technology (SFWST) (formerly called Used Fuel Disposal) Campaign is on repository-induced interactions that may affect the key safety characteristics of EBS bentonite and an argillaceous rock. These include thermal-hydrologicalmechanical- chemical (THMC) process interactions that occur as a result of repository construction and waste emplacement. Some of the key questions addressed in this report include the development of fracturing in the excavation damaged zone (EDZ) and THMC effects on the near-field argillaceous rock and buffer materials and petrophysical characteristics, particularly the impacts of temperature rise caused by waste heat. This report documents the following research activities. Section 2 presents THM model developments and validation, including modeling of underground heater experiments at Mont Terri and Bure underground research laboratories (URLs). The heater experiments modeled are the Mont Terri FE (Full-scale Emplacement) Experiment, conducted as part of the Mont Terri Project, and the TED in heater test conducted in Callovo-Oxfordian claystone (COx) at the Meuse/Haute-Marne (MHM) underground research laboratory in France. The modeling of the TED heater test is one of the Tasks of the DEvelopment of COupled Models and their VAlidation against EXperiments (DECOVALEX)-2019 project. Section 3 presents the development and application of thermal-hydrological-mechanical-chemical (THMC) modeling to evaluate EBS bentonite and argillite rock responses under different temperatures (100 °C and 200 °C). Model results are presented to help to understand the impact of high temperatures on the properties and behavior of bentonite and argillite rock. Eventually the process model will support a robust GDSA model for repository performance assessments. Section 4 presents coupled THMC modeling for an in situ test conducted at Grimsel underground laboratory in Switzerland in the Full-Scale Engineered Barrier Experiment Dismantling Project (FEBEX-DP). The data collected in the test after almost two decades of heating and two dismantling events provide a unique opportunity of validating coupled THMC models and enhancing our understanding of coupled THMC process in EBS bentonite. Section 5 presents a planned large in-situ test, “HotBENT,” at Grimsel Test Site, Switzerland. In this test, bentonite backfilled EBS in granite will be heated up to 200 °C, where the most relevant features of future emplacement conditions can be adequately reproduced. Lawrence Berkeley National Laboratory (LBNL) has very actively participated in the project since the very beginning and have conducted scoping calculations in FY17 to facilitate the final design of the experiment. Section 6 presents present LBNL’s activities for modeling gas migration in clay related to Task A of the international DECOVALEX-2019 project. This is an international collaborative activity in which DOE and LBNL gain access to unique laboratory and field data of gas migration that are studied with numerical modeling to better understand the processes, to improve numerical models that could eventually be applied in the performance assessment for nuclear waste disposal in clay host rocks and bentonite backfill. Section 7 summarizes the main research accomplishments for FY17 and proposes future work activities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Roadmap for disposal of Electrorefiner Salt as Transuranic Waste.

The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

International Approaches for Nuclear Waste Disposal in Geological Formations: Report on Fifth Worldwide Review

An important issue for present and future generations is the final disposal of spent nuclear fuel. Over the past over forty years, the development of technologies to isolate both spent nuclear fuel (SNF) and other high-level nuclear waste (HLW) generated at nuclear power plants and from production of defense materials, and low- and intermediate-level nuclear waste (LILW) in underground rock and sediments has been found to be a challenging undertaking. Finding an appropriate solution for the disposal of nuclear waste is an important issue for protection of the environment and public health, and it is a prerequisite for the future of nuclear power. The purpose of a deep geological repository for nuclear waste is to provide to future generations, protection against any harmful release of radioactive material, even after the memory of the repository may have been lost, and regardless of the technical knowledge of future generations. The results of a wide variety of investigations on the development of technology for radioactive waste isolation from 19 countries were published in the First Worldwide Review in 1991 (Witherspoon, 1991). The results of investigations from 26 countries were published in the Second Worldwide Review in 1996 (Witherspoon, 1996). The results from 32 countries were summarized in the Third Worldwide Review in 2001 (Witherspoon and Bodvarsson, 2001). The last compilation had results from 24 countries assembled in the Fourth Worldwide Review (WWR) on radioactive waste isolation (Witherspoon and Bodvarsson, 2006). Since publication of the last report in 2006, radioactive waste disposal approaches have continued to evolve, and there have been major developments in a number of national geological disposal programs. Significant experience has been obtained both in preparing and reviewing cases for the operational and long-term safety of proposed and operating repositories. Disposal of radioactive waste is a complex issue, not only because of the nature of the waste, but also because of the detailed regulatory structure for dealing with radioactive waste, the variety of stakeholders involved, and (in some cases) the number of regulatory entities involved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Pressurized Slot Testing to Determine Thermo-Mechanical Properties of Lithophysal Tuff at Yucca Mountain Nevada.

The study described in this report involves heated and unheated pressurized slot testing to determine thermo-mechanical properties of the Tptpll (Tertiary, Paintbrush, Topopah Spring Tuff Formation, crystal poor, lower lithophysal) and Tptpul (upper lithophysal) lithostratigraphic units at Yucca Mountain, Nevada. A large volume fraction of the proposed repository at Yucca Mountain may reside in the Tptpll lithostratigraphic unit. This unit is characterized by voids, or lithophysae, which range in size from centimeters to meters, making a field program an effective method of measuring bulk thermal-mechanical rock properties (thermal expansion, rock mass modulus, compressive strength, time-dependent deformation) over a range of temperature and rock conditions. The field tests outlined in this report provide data for the determination of thermo-mechanical properties of this unit. Rock-mass response data collected during this field test will reduce the uncertainty in key thermal-mechanical modeling parameters (rock-mass modulus, strength and thermal expansion) for the Tptpll lithostratigraphic unit, and provide a basis for understanding thermal-mechanical behavior of this unit. The measurements will be used to evaluate numerical models of the thermal-mechanical response of the repository. These numerical models are then used to predict pre- and post-closure repository response. ACKNOWLEDGEMENTS The authors would like to thank David Bronowski, Ronnie Taylor, Ray E. Finley, Cliff Howard, Michael Schuhen (all SNL) and Fred Homuth (LANL) for their work in the planning and implementation of the tests described in this report. This is a reprint of SAND2004-2703, which was originally printed in July 2004. At that time, it was printed for a restricted audience. It has now been approved for unlimited release.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Modeling Radionuclide Releases from TRISO Particles by Simultaneous Diffusion Through and Corrosion of the Silicon Carbide Barrier Layer

TRISO nuclear fuel particles that are less than 1 mm in diameter are designed with multiple barrier layers to retain fission products both during reactor operations and for long-term geological disposal. The primary barrier is a 35 μm thick silicon carbide (SiC-a highly impermeable semi-metal) layer for which data are available on the diffusion of short-lived fission products at high temperatures (> 1000 °C). However, for a geological repository, this layer may contact brine and hence corrode even at ambient temperatures. As an initial approach to assess the effectiveness of the SiC barrier for geological repositories, ranges of fission product diffusivities and corrosion rates for SiC are modeled concurrently with the simultaneous effect of radioactive decay. Using measured corrosion rates of SiC, if the diffusivity is more than about 10 -20 m 2 /s, fission product releases may occur before the SiC barrier has corroded to the point of breach. For diffusivities less than about 10 -21 m 2 /s there may not be significant diffusional releases prior to SiC barrier removal/breach by corrosion. This work shows the importance of estimating diffusivities in SiC at geological repository temperatures, and highlights the relevance of evaluating the porosity/permeability evolution of the SiC layer in a geologic environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Integrating Management of Spent Nuclear Fuel from Generation to Disposal

The current management system in the United States for commercial spent nuclear fuel does not emphasize integration among storage, transportation, and disposal. Unless a path can be implemented that addresses the long-term needs for integration, the United States could end up leaving substantial quantities of stranded commercial spent nuclear fuel stored at decommissioned reactor sites in an increasingly wide variety of containers. This lack of integration does not cause safety issues, but may delay transporting the spent fuel and complicate options for permanent disposal. The large containers now in use for dry storage remain at high temperatures for decades, thereby delaying transportation from decommissioned reactors. The large containers also have no easy path to disposal unless (1) disposal is further delayed (up to 150 years or more for some mined repository concepts); (2) the contents are repackaged into smaller, cooler packages; or (3) the high temperatures are used as de facto site-selection and design criteria for a repository. Implementing consolidated interim storage could address many issues that exist because of this lack of integration. A consolidated interim storage facility that includes appropriate capabilities can allow existing disparate parts to integrate as a system. Previous agencies and commissions have noted this theme before as a way to provide flexibility in the waste management system. This report uses the rationale for such an approach as a framework to discuss the complexities of reconfiguring the waste management system to include consolidated storage. However, concerns that increased storage capacity will reduce the national urgency for a repository are unavoidable, and continued effort will be necessary in public dialogues on the societal aspects of moving commercial spent nuclear fuel into consolidated interim storage. A single optimal solution for integrating current storage and planned transportation with disposal is unlikely. Rather, efforts to integrate various phases of spent fuel management should begin promptly and continue throughout the remaining life of the current fuel cycle. These efforts will need to adapt continuously to evolving circumstances.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Summary of Investigations on Technical Feasibility of Direct Disposal of Dual-Purpose Canisters

This study has evaluated the technical feasibility of direct disposal in a geologic repository, of commercial spent nuclear fuel (SNF) in dual-purpose canisters (DPCs) of existing designs. The authors, representing several national laboratories, considered waste isolation safety, engineering feasibility, thermal management, and postclosure criticality control. The 3-year study concludes that direct disposal is technically feasible for most DPCs, depending on the repository host geology. Postclosure criticality control, and thermal management strategies that allow permanent disposal within 150 years, are two of the most challenging aspects. This document summarizes technical results from a series of previous reports, and describes additional studies that can be done especially if site-specific information becomes available from one or more prospective repository sites.

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

Sustainable Data Evolution Technology (SDET) for Power Grid Optimization (Final Report)

This project develops an innovative sustainable data evolution technology (SDET) to create open-access power grid datasets and facilitate updates to these datasets by the power grid community. The lack of open-access, realistic large-scale datasets significantly limits the ability of researchers, developers and ultimately end users to develop, benchmark and compare new methods and tools for optimizing the operation and planning of the grid, which leads to slow adoption by end users. Available open datasets include the IEEE sponsored 14-300 bus transmission test cases and distribution test feeders, the Edinburgh Power Systems Test Case Archive (39-2224 buses), and the Polish cases (2383-3375 buses) available with MATPOWER. These datasets serve certain purposes, but for power grid optimization, they are insufficient in one or more of these aspects: 1) they are too small; 2) they are incomplete in both models and scenarios for optimization purposes; and 3) they are static and are not keeping up with grid needs. The SDET technology delivers large-scale realistic datasets and data-creation tools capable of generating new datasets. We work closely with the Category-2 data repository project awarded by the same GRID DATA program. The generated datasets and data creation tools are compatible with and available through the data repository. The objective is to make this a sustained effort within and beyond the ARPA-E GRID DATA program so that the datasets can evolve over time and meet the current and future needs for power grid optimization and potentially other applications in power grid operation and planning. The novel and disruptive features of SDET differentiate itself from prior work in the following aspects: 1) SDET derives features and metrics for both transmission and distribution (T&D) systems by analyzing many public and private datasets provided by our industry partners National Rural Electric Cooperative Association (NRECA) and PJM Interconnection (PJM). 2) SDET develops data-creation tools and use these tools to generate large-scale open-access realistic datasets that comply with the metrics for both T&D systems. 3) SDET validates the created datasets using industry tools provided by our vendor partner GE Energy Solution. The data creation tools, as well as the created datasets, are integrated into the data repository (DR. POWER) funded by ARPA-E GRID DATA program in Category 2. Users can use the datasets and access these data creation tools to create and update datasets, thus datasets can evolve per user requirements and power grid development. The SDET is a novel and disruptive technology compared to current ad hoc and static dataset generation. It is expected the SDET technology will fundamentally enhance the development, assessment, and adoption of new methods and tools for power grid optimization and other applications, which in turn would significantly improve the reliability, resiliency, and efficiency of the power grid – a major ARPA-E mission objective.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2019

DOE SFWST Campaign R&D Roadmap Update Rev.1.

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Departmentof Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel and Waste Disposition(SFWD) is conducting research and development (R&D) on deep geologic disposal of spentnuclear fuel (SNF) and high-level nuclear waste (HLW). R&D addressing the disposal ofSNF/HLW in the U.S. is currently generic (i.e., "non-site-specific") in scope, following thesuspension of the Yucca Mountain Repository Project in 2010. However, to prepare for theeventuality of a repository siting process, the former Used Fuel Disposition Campaign (UFDC) ofDOE-NE, which was succeeded by the SFWST Campaign, formulated an R&D Roadmap in 2012outlining generic R&D activities and their priorities appropriate for developing safety cases andassociated performance assessment (PA) models for generic deep geologic repositories in severalpotential host-rock environments in the contiguous United States. This 2012 UFDC Roadmap alsoidentified the importance of re-evaluating priorities in future years as knowledge is gained fromthe DOE's ongoing R&D activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Innovative Linear and Nonlinear Solvers for Simulating Multiphase Flow within Large-Scale Engineered Subsurface Systems

Numerical simulation of non-isothermal multiphase porous flow combined with reactive transport is used in a wide range of applications, which include nuclear waste repositories, enhanced recovery of petroleum reservoirs, contaminant remediation, geothermal engineering, and carbon sequestration. Understanding and predicting underground phenomena can have enormous impact on how to deal with world issues like climate change, clean water, and renewable energy. The main motivation for this proposal arises from safety assessment of future nuclear waste repositories using the U.S. Department of Energy (DOE) Geologic Disposal Safety Assessment (GDSA) Framework, and performance assessment (PA) for Waste Isolation Pilot Plant (WIPP), the nation's only active nuclear waste repository in Carlsbad, New Mexico. I am a technical staff member at Sandia National Laboratories involved in both projects.

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

Cost Estimation Inputs for Spent Nuclear Fuel Geologic Disposal Concepts

A set of 16 geologic disposal concepts is described in sufficient detail for rough-order-of-magnitude repository cost estimates, for disposal of spent nuclear reactor fuel in generic crystalline, argillaceous, and salt host geologic media. The description includes total length, diameter, and volume for all underground shafts, ramps, drifts and large-diameter borings. Basic types of ground support are specified. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel, but concepts are described in terms of modular panels each containing 10,000 MT. Waste packaging is described, and the materials and outer dimensions for disposal overpacks are given. The manner of emplacement is specified, with any additional fixturing, lining, buffer materials, and backfill needed. Thermal limits for waste package emplacement or repository closure are given, as appropriate for enclosed and open emplacement modes, respectively. The 16 disposal concepts are based on disposal concept studies performed for the Used Fuel Disposition R&D program between 2011 and 2014, and the accompanying engineering analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Benched: Upgrading and Updating the Los Alamos Benchmark Suite for the 21st Century [Slides]

The presentation begins by stating that most organizations make choices to handle specific things when preparing input files for validation, such as which elemental abundance data to use, how to decompose specific elements, and setting calculation precision. It also discusses the consequences, which include differing calculation results between different organizations for the same case, even when using the same code. This is illustrated by the intercomparison exercise and input files that are prepared by one organization and are not necessarily correct for another. The presentation explains that the Los Alamos Benchmark Suite is a centralized repository of high-quality benchmark models currently under development at LANL. The objectives of the new repository are to remain up to date with the latest ICSBEP revision, to have a formal review and revision process, to (eventually) be an open-source repository, and to provide new tools for improved input and output file generation and review.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

International Collaboration Activities on Engineered Barrier Systems

International collaborations on nuclear waste disposal is an integral part of the Spent Fuel Waste Science and Technology (SFWST) campaign within the DOE Fuel Cycle and Technology (FCT) program. These engagements with international repository R&D programs provide key opportunities to participate in experiments with international partners on research investigations developing laboratory/field (underground research laboratories (URL) experiments) data of engineered barrier system (EBS) components (e.g., near-field) and characterization of transport phenomena in the host rock (e.g., far-field). The results of these field and laboratory experiments are used in the evaluation of coupled processes and the development of state-of-the-art simulation approaches to evaluate repository performance. Thermal heating from radionuclide decay in the waste canisters will increase temperature in the surrounding EBS driving chemical and transport processes in the near- and far-field domains of the repository. URL heater-tests for extended periods of times (e.g., years) provide key information and data of thermal effects on barrier responses to temperature and water saturation levels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Preliminary Engineering and Cost Analysis for DPC Disposal Solutions. Final Draft

There are currently (as of January, 2019) more than 2,700 dual-purpose canisters (DPCs) loaded with spent nuclear fuel (SNF) across the United States. DPCs continue to be loaded at a rate of more than 200 per year by mid-century there are likely to be more than 8,160 DPCs in service. Options for disposing of SNF loaded in DPCs include repackaging into specialized disposal canisters, directly disposing of the loaded DPCs (with or without modification), or some combination of the two. The main technical challenges for direct disposal of loaded DPCs are thermal management, handling and emplacement operations for the large, heavy packages, and postclosure criticality control. This report focuses on postclosure criticality control which is the most challenging. The challenge lies in determining how to modify DPCs so as to minimize the probability that a criticality event might occur in a repository, or if the DPCs are not modified, to understand the nature and consequences of postclosure criticality events. There are several approaches that could facilitate direct disposal of loaded DPCs with acceptable repository performance. This report describes these approaches and presents comparative analysis of the rough-order-of-magnitude (ROM) costs. Repackaging SNF in DPCs into specialized disposal canisters could be financially and operationally costly with additional radiological, operational safety, and management risks. A disposition approach that would not involve repackaging or modifications to DPCs (future or already loaded) is development of a new licensing strategy that addresses the risk (probability and consequence) from criticality events. A different approach would modify existing loaded DPCs (some or all of them), and change the loading or design of future DPCs, to decrease the probability of a criticality event in a repository below levels of concern. This report investigates the cost to modify existing loaded DPCs, and the cost to modify the loading or design of future DPCs to facilitate direct disposal. It establishes the ROM cost for repackaging SNF that has been loaded into DPCs, into specialized canisters for disposal. It also identifies technical and regulatory challenges associated with the potential design modifications and loading considerations. It is left to future analyses to compare radiological, operational safety, and management risks associated with the available approaches.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

DOE SFWST Campaign R&D Roadmap Update

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel and Waste Disposition (SFWD) is conducting research and development (R&D) on deep geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). R&D addressing the disposal of SNF/HLW in the U.S. is currently generic (i.e., "non-site-specific") in scope, following the suspension of the Yucca Mountain Repository Project in 2010. However, to prepare for the eventuality of a repository siting process, the former Used Fuel Disposition Campaign (UFDC) of DOE-NE, which was succeeded by the SFWST Campaign, formulated an R&D Roadmap in 2012 outlining generic R&D activities and their priorities appropriate for developing safety cases and associated performance assessment (PA) models for generic deep geologic repositories in several potential host-rock environments in the contiguous United States. This 2012 UFDC Roadmap also identified the importance of re-evaluating priorities in future years as knowledge is gained from the DOE's ongoing R&D activities.

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

Dual Purpose Canister Reactivity and Groundwater Absorption Analyses

The current spent nuclear fuel (SNF) management strategy includes reliance on dry storage. Utilities are meeting their interim storage needs on an individual basis with use of large-capacity dry storage casks, with a current focus on meeting existing storage and transportation requirements, as disposal requirements are not currently available. These casks are commonly known as dual-purpose (i.e., storage and transportation) canisters (DPCs). However, a small percentage of single-purpose (storage only) systems is also being used to meet storage needs. These are included under the “DPC” heading. This report investigates the postclosure criticality safety aspects of DPCs. Placing large, heavy waste packages containing DPCs into a repository for direct disposal has not yet been implemented domestically or internationally. Therefore, direct disposal of DPCs represents new engineering and scientific challenges. Some of the engineering challenges that have already been addressed include handling and placement, use of ramps vs. shafts, use of hoists, use of transport equipment, and thermal management. Additionally, some studies have been conducted in the past regarding the feasibility of direct disposal from a criticality analysis perspective and have concluded that while possible, demonstrating subcriticality over the disposal time period is a challenge. The alternative to direct disposal of DPCs into a repository is to repackage the SNF into different canisters. The direct disposal of DPCs without cutting them open and repackaging is appealing because it could be more cost-effective, reduce the complexity of fuel management operations both in and outside reactor facilities, and result in less cumulative worker dose during interim storage and handling before eventual disposal in a deep geologic repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Overview of Advanced Reactor Simulation Capabilities to Support Demonstrations

The Virtual Test Bed (VTB) was established to support the deployment of advanced reactors by facilitating the use of state-of-the-art modeling and simulation capabilities. These tools are able to solve previously prohibitive challenge problems in greater levels of fidelity and in a multi-physics framework. A VTB repository was recently set up to host models for a wide variety of reactor types and using a range of NEAMS-developed software. Table I summarizes the different capabilities already available in the repository and up-and-coming near/long term ones. The repository currently hosts models for a wide variety of reactor types, including High-Temperature Gas Cooled Reactors (HTGR), Sodium Fast Reactors (SFR), Fluoride salt cooled High-temperature Reactors (FHR), Molten Salt Reactors (MSR), and Heat-Pipe Microreactors (HP-MR). It should be noted that work is ongoing for all of the models and each example problem is at a different state of maturity.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

Status Update on the High Precision Isotopic Measurements on High Burnup LWR Fuel in 2020

The US Department of Energy (DOE) Office of Nuclear Energy (NE) is currently investigating the feasibility of directly disposing dual-purpose (storage and transportation) canisters (DPCs) in a spent nuclear fuel (SNF) repository. Criticality during the repository performance period (10,000 years or more) is one of the major concerns related to direct disposal of DPCs, specifically as the system undergoes degradation in the repository environment and timeframe. Oak Ridge National Laboratory (ORNL) is developing an as-loaded criticality analysis methodology using full (actinides + fission products) burnup credit that exploits the inherent criticality margin associated with actual canister-specific loading configuration. Burnup credit criticality analysis requires validation of the depletion/decay codes used to generate the burned isotopic inventory of an assembly by comparing the code-predicted inventory with experimentally determined isotopic data. Currently, isotopic measurement data for boiling water reactors (BWR) SNF are limited, and additional measurements will be highly beneficial for BWR burnup credit analysis, which is essential to demonstrate disposability of BWR DPCs. Moreover, new pressurized water reactor (PWR) samples of isotopic measurements will expand the PWR sample population and consequently will greatly improve sample statistics. In turn, this will reduce uncertainty in the computational determination of the isotopic composition of commercial SNF and eliminate additional penalties currently used for lack of data. As part of the isotopic analysis task, eight diverse, high burnup (HBU) samples from several PWR rods are being dissolved and isotopically analyzed to provide high-quality measurement data for PWR SNF to reducing the uncertainties associated with PWR isotopic depletion validation. The isotopics of interest focus on nuclides important to burnup credit, shielding, and decay heat in PWR SNF. Additionally, eight high burnup BWR fuel samples from the Limerick nuclear power plant have been identified for measurement to support reducing uncertainties associated with BWR isotopic depletion validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗