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At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Comparison of Light-Weight Radioisotope Heater Unit (LWRHU) Historical Testing to Special Form Testing Requirements

This paper compares the Light-Weight Radioisotope Heater Unit (LWRHU) historical testing to Department of Transportation (DOT) Special Form criteria per 49 CFR Part 173, Shippers-General Requirements for Shipments and Packages, Section 469, Tests for special form Class 7 (radioactive) materials as specified in DOE-STD-1027, Hazard Categorization of DOE Nuclear Facilities. Compliance with the aforementioned criteria allows for the exclusion of radioactive sources from the facility’s nuclear material inventory. As part of a new program at Los Alamos National Laboratory (LANL), LWRHUs are planned to be moved into TA-55, PF-5 Power Source Assembly Area (PSAA) for assembly and testing activities. The PSAA Facility is categorized as a low hazard, non-nuclear facility per DOE STD-1027 and is described in TA55-PF-5-FHC-11-001, Facility Hazard Categorization of the PSAA at PF-5, TA-55. The Facility Hazard Categorization document states, there must be documentation that radioactive sources are engineered to pass the special form testing specified by DOT in 49 CFR 173.469, Tests for special form Class 7 (radioactive) materials or ANSI N43.6, Sealed Radioactive Sources - Classification. This paper documents the extensive historical environmental safety testing of the LWRHU and demonstrates its ability to survive potential launch safety accident scenarios. The testing results are used herein to demonstrate through a comparative analysis that the robust design of the LWRHU meets or exceeds the special form testing specifications in 49 CFR 173.469, enabling the LWRHU to be transported and handled in PF-5 in support of testing and assembly program.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Transportation and Systems Analysis Collaborations in Support of a Federal Consolidated Interim Storage Facility

The U.S. Department of Energy’s Integrated Waste Management (IWM) program under the Office of Nuclear Energy is planning for the future transportation, storage, and eventual disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from nuclear power plant sites across the United States. To better enable informed decision-making regarding the back end of the nuclear fuel cycle, the IWM program has been sponsoring the development and application of system analysis tools capable of analyzing various options for managing SNF and HLW. With these tools, integrated waste management system (IWMS) architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). System analyses and assessments typically use these modeling and simulation tools to investigate implications of changes in various assumptions and parameters such as acceptance rates, receipt logic, facility capacities and capabilities, use of standardized canisters, start and stop dates of facilities, etc. The Next Generation System Analysis Model (NGSAM) is an agent-based simulation toolkit that is used for a system-level simulation and analysis focused on SNF management in the United States. An analyst using NGSAM has the ability to define several factors like the number of storage facilities, capacity at each facility, transportation schedules, shipment rates, and other conditions. NGSAM’s primary purpose is to provide a system analyst with capabilities to model the IWMS and gain insights into SNF and HLW management alternatives including the impact of system choices, associated cost estimates, and development of integrated yet flexible approaches. IWM is also developing the Stakeholder Tool for Assessing Radioactive Transportation (START). START is a web-based geospatial tool developed to provide visualization and initial evaluation of transportation options associated with future SNF and HLW shipment planning and operations. This includes characterizing safety, economic, and environmental conditions on and in proximity to shipment origins as well as along prospective transportation routes. Information from the START tool can be presented/shared as maps, graphics, geospatial files, and tabular form to enable easy data representation and export functionality. The system analysis, NGSAM, and START teams had been working closely for several years before formalizing this collaboration. START provides the SNF routing information for use in NGSAM. System analysts use the NGSAM tool to generate results (schedule, costs, infrastructure acquisitions, shipment rates, etc.). Subject-matter experts and system analysts work together to inform how NGSAM should model the waste management system. Specifically, this paper discusses the collaborations between the system analysis, NGSAM, and START teams and their accomplishments over the past year. Some examples of these efforts include calibrating and adding data to the START output files to meet NGSAM needs, gaining a better understanding of START data used in NGSAM, as well as how updates in START data could result in an improved NGSAM analysis. Detailed examples of various tasks that have been performed by the team will be discussed in the full paper. Continued efforts in this direction are expected in the coming years.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Defect-induced phonon-resonant scattering and its influence on thermal transport of irradiated thorium-dioxide

Thermal transport in proton irradiated thorium-dioxide (ThO 2 ) is investigated. Using a combination of experiments and first-principles computational framework, the role of lattice defects on thermal conductivity is analyzed. A resonant-phonon scattering mechanism beyond the traditionally considered Rayleigh scattering is found to significantly influence low-temperature thermal transport in the presence of irradiation-induced point defects. The existence of localized phonon modes associated with irradiation-induced defects is suggested by the inability of the first-principles based thermal conductivity model—which considers only three-phonon interactions and phonon-defects scattering using the Tamura formalism—to predict the experimental results, unless a resonant scattering mechanism is included. The emergence of additional peaks in the Raman spectra in the proximity of phonon-resonant frequency provides further evidence for the existence of localized modes. Coupled with a microstructure evolution model, this analysis enables more accurate analysis for contrasting the contributions of different phonon scattering mechanisms across all irradiation doses and temperatures.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2005

An Algorithm for the Transport of Anisotropic Neutrons

One major obstacle to human space exploration is the possible limitations imposed by the adverse effect of long-term exposure to the space environment. Even before human spaceflight began, the potentially brief exposure of astronauts to the very intense random solar particle events (SPE) were of great concern. A new challenge appears in deep space exploration from exposure to the low-intensity heavy-ion flux of the galactic cosmic rays (GCR) since the missions are of long duration and the accumulated GCR exposures can be high. Because cancer induction rates increase behind low to rather large thicknesses of aluminum shielding, according to available biological data on mammalian exposures to GCR like ions, the shield requirements for a Mars mission are prohibitively expensive in terms of mission launch costs. Therefore, a critical issue in the Human Exploration and Development of Space enterprise is cost effective mitigation of risk associated with ionizing radiation exposure. In order to estimate astronaut risk to GCR exposure and associated cancer risks and health hazards, it is necessary to do shield material studies. To determine an optimum radiation shield material it is necessary to understand nuclear interaction processes such as fragmentation and secondary particle production which is a function of energy dependent cross sections. This requires knowledge of material transmission characteristics either through laboratory testing or improved theoretical modeling. Here ion beam transport theory is of importance in that testing of materials in the laboratory environment generated by particle accelerators is a necessary step in materials development and evaluation for space use. The approximations used in solving the Boltzmann transport equation for the space setting are often not sufficient for laboratory work and those issues are a major emphasis of the present work.

Tweed, J.↗
DOE OSTI2024

Considerations for Managing DOE Standard Canisters within an Over-canister as Part of an Integrated Waste Management System PPT

To better enable informed decision making regarding the back-end of the nuclear fuel cycle, the Integrated Waste Management Program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring research into a comprehensive integrated waste management system (IWMS) that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). An important aspect of the IWMS is DOE-managed Spent Nuclear Fuel (SNF). DOE and its predecessor agencies have generated, transported, received, stored, and reprocessed SNF at DOE facilities nationwide, and DOE is responsible for managing the SNF currently in its possession. These fuels come from a wide range of reactor types that employ various cladding materials, fuel materials, and enrichments. To enable interim, road-ready dry storage (RRDS) of the wide variety of SNF types found in the DOE inventory, a standardized canister system (i.e., the DOE Standard Canister) was proposed for the packaging demonstration. This robust, welded canister system is designed to confine radionuclides, prevent criticality by precluding content moderation, and satisfy other requirements as part of a larger storage, transportation, and disposal system. While SNF has yet to be loaded into a DOE Standard Canister, DOE Standard Canister designs were included in past storage facility and disposal facility design licensing endeavors. In a renewed effort to evaluate packaging SNF at Idaho National Laboratory (INL) in a RRDS configuration, researchers are planning the RRDS Packaging Demonstration. This demonstration is supplemented by analytical structural, criticality, and material compatibility evaluations that support management of SNF in DOE Standard Canisters, taking advantage of past analysis work to the extent possible. One of the largest differences between the current Packaging Demonstration and past analytical evaluations is the inclusion of an over-canister containing multiple DOE Standard Canisters. For the Packaging Demonstration, DOE Standard Canisters loaded with SNF are planned to be placed in a larger diameter over-canister. The sealed over-canister could then be placed in a storage overpack for onsite storage, or in a transportation overpack for shipment to an offsite storage location or disposal site once one becomes available. This paper examines the relevant considerations and provides a preliminary evaluation of integrating the over-canister configuration into the storage, transportation, and disposal processes of the overall waste management system. For storage and transportation, the over-canister can be considered analogous to a multi-purpose canister (MPC) for commercial SNF. For disposal, the DOE Standard Canisters could be removed from the over-canister and placed in a co-disposal waste package with canisters containing vitrified high-level radioactive waste (HLW) similar to configurations examined previously, or the sealed over-canisters might be capable of direct disposal in a waste package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

CRISPR–Cas9-mediated nuclear transport and genomic integration of nanostructured genes in human primary cells

DNA nanostructures are a promising tool to deliver molecular payloads to cells. DNA origami structures, where long single-stranded DNA is folded into a compact nanostructure, present an attractive approach to package genes; however, effective delivery of genetic material into cell nuclei has remained a critical challenge. Here, we describe the use of DNA nanostructures encoding an intact human gene and a fluorescent protein encoding gene as compact templates for gene integration by CRISPR-mediated homology-directed repair (HDR). Our design includes CRISPR–Cas9 ribonucleoprotein binding sites on DNA nanostructures to increase shuttling into the nucleus. We demonstrate efficient shuttling and genomic integration of DNA nanostructures using transfection and electroporation. These nanostructured templates display lower toxicity and higher insertion efficiency compared to unstructured double-stranded DNA templates in human primary cells. Furthermore, our study validates virus-like particles as an efficient method of DNA nanostructure delivery, opening the possibility of delivering nanostructures in vivo to specific cell types. Together, these results provide new approaches to gene delivery with DNA nanostructures and establish their use as HDR templates, exploiting both their design features and their ability to encode genetic information. This work also opens a door to translate other DNA nanodevice functions, such as biosensing, into cell nuclei.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2025

Local aging effects in PuB 4 : Growing inhomogeneity and slow dynamics of local field fluctuations probed by 239 Pu NMR

Plutonium-based correlated electron materials host exotic physical phenomena ranging from unconventional heavy-fermion superconductivity to topological Kondo insulating states. Self-irradiation damage can influence many properties of such radioactive materials. Structural disorder effects due to α radiation have been frequently studied using techniques such as transport, thermodynamics, and x-ray diffraction. Here, in this study, we use 239 Pu nuclear magnetic resonance (NMR) to study the long-term influence of self-damage on the lattice and local electronic structures in a single crystal of the candidate topological insulator plutonium tetraboride (PuB 4 ). We first characterize the anisotropy of the 239 Pu resonance and confirm the local axial-site symmetry inferred from previous polycrystalline measurements. Aging effects are then evaluated over the time frame of six years. We find that, though the static 239 Pu NMR spectra show a slight modulation in their shape, their field-rotation pattern reveals no change in 239 Pu local site symmetry over time, suggesting that aging has a surprisingly small impact on the spatial distribution of the static hyperfine field. Further, ligand-site 11 B NMR finds little time-dependent change in the size of electric field gradient around 11 B sites. By contrast, aging has a prominent impact on the 239 Pu NMR relaxation processes and signal intensity. Specifically, aging-induced damage manifests itself as an increase in the spin-lattice relaxation time 𝑇 1 , an increased distribution of 𝑇 1 , and a signal intensity that decreases linearly by 20% per year. An effective spin-spin relaxation time 𝑇 2,eff in the aged sample shortens drastically towards lower temperature, suggesting growth of slow fluctuations of the hyperfine field that are linked to radiation-damage-induced inhomogeneity. Our NMR study sheds light on the interplay of radiation damage and local magnetic interactions in correlated insulators.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2020

VARI3D & PERSENT: Perturbation and Sensitivity Analysis

The nodal diffusion method is one of the most widely used approaches in modern reactor analysis. In the nodal diffusion method, a coarse multi-group set of “homogenized” parameters is constructed such that the complex geometry of a reactor core along with the energy dependence of neutron and gamma ray cross sections in a nuclear reactor are conserved in the simpler geometry. The homogenization is typically done on a fuel assembly level as is the case in the DIF3D code developed at Argonne National Laboratory. The nodal methodology is used primarily to predict fuel cycle behavior of nuclear systems of which there is a substantial amount of validation in the literature. Another use of the nodal method is to obtain reactivity coefficients and kinetics parameters for use in a safety analysis of a given nuclear reactor. While there are many ways to obtain reactivity worth and kinetics parameters, the work presented in this manuscript is unique as it provides the user with the ability to compute reactivity worths, kinetics parameters, and cross section sensitivities with a Cartesian and hexagonal geometry based transport code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

VARI3D & PERSENT: Perturbation and Sensitivity Analysis

The nodal diffusion method is one of the most widely used approaches in modern reactor analysis. In the nodal diffusion method, a coarse multi-group set of “homogenized” parameters is constructed such that the complex geometry of a reactor core along with the energy dependence of neutron and gamma ray cross sections in a nuclear reactor are conserved in the simpler geometry. The homogenization is typically done on a fuel assembly level as is the case in the DIF3D code developed at Argonne National Laboratory. The nodal methodology is used primarily to predict fuel cycle behavior of nuclear systems of which there is a substantial amount of validation in the literature. Another use of the nodal method is to obtain reactivity coefficients and kinetics parameters for use in a safety analysis of a given nuclear reactor. While there are many ways to obtain reactivity worth and kinetics parameters, the work presented in this manuscript is unique as it provides the user with the ability to compute reactivity worths, kinetics parameters, and cross section sensitivities with a Cartesian and hexagonal geometry based transport code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Designing Flexibility into the Authorized Radioactive Contents of the 9982 Shipping Package

The 9982 radioactive material shipping package, or the 9982, differs from the others by allowing a multitude of radionuclides instead of a single defined set. While allowing flexibility, this requires more justification on why these contents would be worthwhile and why they would not cause various safety hazards when shipped together. The latter was covered through mass limit calculations, and the potential uses for the package are shown through comparison with historical shipments. Here, the mass limits were already calculated[1] and used to show three example shipments are suitable for the 9982 as well. Plutonium stored in 3013 containers[2], training sources shipped in 9977 packages[3], and americium standards shipped in 9978 packages[4] were all analyzed. The mass of each radionuclide and the corresponding mass limit were used to first determine if the listed content could be shipped. If so, the maximum allowed mass of such contents is also calculated in order to see if the 200-gram fissile limit for the 9982 would be violated if additional material was packed. The sum-of-the-fractions method produces contents that do not violate the mass limits or exceed the 200-gram fissile limit. These sets also serve as realistic examples of what may be shipped in the 9982. Well-founded example contents will help the approval of the package by showing its compatibility with previously shipped material. Unfounded hypothetical contents would not show the usefulness of having a wide variety of allowed radionuclides. More contents may be developed in order to incorporate other radionuclides that were not present in these three examples. The light elements will also need to be evaluated due to their exclusion currently preventing some materials from being allowed in the 9982.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2005

How Space Radiation Risk from Galactic Cosmic Rays at the International Space Station Relates to Nuclear Cross Sections

Space radiation risk to astronauts is a major obstacle for long term human space explorations. Space radiation transport codes have thus been developed to evaluate radiation effects at the International Space Station (ISS) and in missions to the Moon or Mars. We study how nuclear fragmentation processes in such radiation transport affect predictions on the radiation risk from galactic cosmic rays. Taking into account effects of the geomagnetic field on the cosmic ray spectra, we investigate the effects of fragmentation cross sections at different energies on the radiation risk (represented by dose-equivalent) from galactic cosmic rays behind typical spacecraft materials. These results tell us how the radiation risk at the ISS is related to nuclear cross sections at different energies, and consequently how to most efficiently reduce the physical uncertainty in our predictions on the radiation risk at the ISS.

Lin, Zi-Wei↗
DOE OSTI2025

SULI Intern Final Report: Computationally Investigating Hydrogen Thermo-Diffusion in Yttrium Hydride Using Multiscale Methods

The renaissance of nuclear energy has arrived, heralding an age of abundant inexpensive clean energy, and renewed space exploration. In nuclear-powered spacecraft and microreactors, safety and size are of utmost importance. Yttrium Hydride (YHx) is being researched for its utility as a neutron moderator in nuclear reactors; the hydrogen in YHx slows down neutrons, enabling a continuous nuclear reaction in the reactor. This has the benefit of allowing reactors to be more safe, compact, and efficient. The goal of this effort is to computationally predict the coefficient of temperature-dependent hydrogen diffusion within YHx, the Soret coefficient. This parameter is essential for determining the safe operating modes of YHx moderators. Zirconium Hydride (ZrHx) is used in the Training, Research, Isotopes, General Atomics (TRIGA) reactor, is the reference material for these calculations. In this work, nanoscale atomic modeling in the Vienna Ab initio Simulation Package (VASP) is combined with the mesoscale finite element phase-field module in the Multiphysics Object-Oriented Simulation Environment (MOOSE); this culminates in a new multiscale computational method to simulate Soret diffusion of hydrogen in YHx. This data is useful for predicting experimental outcomes. This workflow involves convergence testing followed by static, Nudged Elastic Band (NEB), Quasi-Harmonic Approximation (QHA), and Molecular Dynamics (MD) calculations - linked with phase field simulation. NEB simulates hydrogen migration, while QHA and MD predict temperature-dependent properties. The static calculations align with literature, and preliminary NEB and QHA calculations yield accurate results. Once the atomic calculations are complete, we will incorporate Electron Backscatter Diffraction (EBSD) images and VASP-generated parameters into the phase field module to simulate intra- and intergranular transport of hydrogen in ZrHx and YHx. Future research will extend our approach to fuel-moderator materials systems such as Uranium-Yttrium Hydride (U-YHx). This work contributes to the development of advanced nuclear energy solutions for space travel.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Spent nuclear fuel receipt rate analysis within an integrated waste management system (IWMS) architecture that includes consolidated storage

A key parameter in analyzing the performance of an integrated waste management system (IWMS) architecture for the disposition of spent nuclear fuel (SNF) is the SNF receipt rate from reactor and other custodian sites. Receipt rate in this paper means how much SNF is accepted per year for transport in the IWMS from such sites. Introducing one or more federal consolidated interim storage facilities (CISFs) into the IWMS architecture can potentially accelerate the receipt rate profile over time relative to system architectures without a CISF. This raises the question of what an optimal SNF receipt rate profile for an IWMS architecture might be in view of practical constraints and desired system performance attributes and associated metrics. This paper describes a sensitivity study on SNF receipt rates and the associated results for a selected set of IWMS scenarios aimed at informing near-term planning for interim storage capabilities and transportation assets. Two different strategies for CISF operation while awaiting availability of a disposal system to receive SNF are compared: one that relatively quickly fills an initial CISF and then idles the transportation system; and another that aims for more continuous use of transportation assets and receipt capabilities at the CISF. This study examines cost considerations and other factors, such as the timing of clearing reactor sites of SNF, efficient use of capital assets, and some other metrics that might be important to a CISF host community. Based on the analysis, an initial approach is presented that targets a continuous receipt strategy while maintaining the flexibility to step up receipt capabilities to a reasonable degree when needed and beneficial, within overall system constraints.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Development of the 3 m{sup 3} Low Level Radioactive Waste Container in Taiwan - 20037

Chinshan Nuclear Power Plant stopped its business operation in July, 2019, and Kuosheng Nuclear Power Plant will also be permanently shut down in 2023. Nuclear Reactor Facilities Regulation Act requires a nuclear facility must be dismantled in 25 years after being permanently ceased operation. According to the approved decommissioning plan of Chinshan Nuclear Power Plant and the proposed one of Kuosheng Nuclear Power Plant, the 3 m{sup 3} low level radioactive waste container is planned to store the low level radioactive wastes with higher radiation produced during the decommissioning. However, 3 m{sup 3} low level radioactive waste containers are not used or developed in Taiwan before. To meet such need, Institute of Nuclear Energy Research develops a new type of 3 m{sup 3} low level radioactive waste container and performs the corresponding analyses and tests based on Guidelines of Applying the Usage License of Low Level Radioactive Waste Containers and Regulations for the Safe Transport of Radioactive Material. In this article, it provides the information of the low level radioactive waste containers that have been approved or is under reviewing in Taiwan, and so do the design specifications of the developed 3 m{sup 3} waste container. Besides, the numerical evaluation and test results of the developed 3 m{sup 3} waste container are also included in this article. Based on the evaluation and test results, the developed 3 m{sup 3} waste container satisfies the requirements of the storage container and industrial package type 2. Institute of Nuclear Energy Research will submit the usage license applications to the authority this year, and the usage licenses would be granted in 2021, expectedly. (authors)

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

New Developments and Verification of the Fusion Blanket Simulation Capabilities in the MOOSE Framework

This poster contains new developments and updates in the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework related to fusion modeling and simulation, and forms the basis for an on-going effort in modeling whole fusion devices at Idaho National Laboratory using MOOSE. In this work, a slice of the midplane of a proposed ceramic breeder blanket design for the Fusion Nuclear Science Facility is simulated using a combination of neutronics, 1D system level thermal hydraulics for helium, 3D heat conduction, and tritium transport. Previous work focused on the engineering scale full assembly of a blanket sector, but the simulations here form the basis for software quality and verification efforts to increase confidence as the large scale simulations continue. This poster was presented at the Idaho National Laboratory Intern Poster Session 2023.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Outcomes of WPEC SG47 on "Use of Shielding Integral Benchmark Archive and Database for Nuclear Data Validation"

The Working Party on International Nuclear Data Evaluation Co-operation Subgroup 47 (WPECSG47) entitled "Use of Shielding Integral Benchmark Archive and Database for Nuclear Data Validation" was organised from 2019 and 2022 with the objectives to promote more systematic and wider use of shielding benchmark experiments in nuclear data (ND) and transport code validation and development, to provide feedback on the Shielding Integral Benchmark Archive and Database (SINBAD), and to promote its further development in coordination with the Expert Group on Physics of Reactor Systems (EGPRS). Altogether 9 meetings, the large majority (8) held remotely, were organised during the past 3 years to discuss the experience on the use of SINBAD, evaluation of new benchmarks and improvements to be contributed to the database which was severely neglected and lacking maintenance over the past ← 10+ years. Several proposals for new or updated benchmark evaluation were presented and discussed, such as FNG copper, LLNL pulsed spheres, CIAE iron sphere, KFK 1977 gamma measurements, Rez Fe sphere, ASPIS, ORNL Oxygen broomstick, TIARA and others. Complementing the database with new features was also discussed, for example providing the nuclear data sensitivity profiles more systematically would facilitate and better guide the use of data. Information on the geometry, (radiation source) and materials available in CAD format is expected to allow an easier and less error prone reference for computational model preparation and a potential input to CAD based workflows. Inputs for various transport codes and other benchmark data from participants have been shared via the NEA GitLab which could hopefully in the future evolve and form a bases for critically checked and validated benchmark data. Future development of SINBAD will be monitored by EGPRS and the newly created SINBAD Task Force.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Tritium Transport Modeling at the Pore Scale in Ceramic Breeder Materials Using TMAP8

Fusion reactors depend on the blanket material to breed and release tritium at the same rate or faster than it is consumed by the fusion reaction. Cellular ceramic breeders (CCBs) are dense materials that can maintain a high tritium breeding ratio while promoting tritium release because of highly connected pores. Assessing the tritium breeding capabilities of these materials requires a combination of extensive experimental and modeling efforts. In this work, we develop and calibrate a multiphysics model of tritium transport. This novel model accounts for ceramic and pore diffusion, trapping and detrapping, and several surface reactions at the pore surface. We perform a sensitivity analysis and calibrate the model by comparing its predictions against experimental measurements of deuterium absorption. The calibrated model is then used to model tritium absorption in samples with different pore microstructures to investigate the effect of pore interconnectivity on tritium absorption. The model is part of the development of the multiscale, multiphysics framework for tritium transport [i.e., the Tritium Migration Analysis Program (TMAP8)], which is itself built on top of the finite-element multiphysics framework multiphysics object-oriented simulation environment (MOOSE). This study demonstrates some of TMAP8’s capabilities and is the first step toward assessing the tritium breeding capabilities of ceramic breeder material designs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗