Advanced Reactor Development: Gas-Cooled Reactor First Quarter Report
Significant and major accomplishments of the Advanced Reactor Development: Gas-Cooled Reactor Program for the first quarter of 2021.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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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.
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Four Design Demonstration Elments (DDEs) to be irradiated in the Advanced Test Reactor (ATR) and the Belgian Reactor 2 (BR2)
The U.S. High Performance Research Reactor Conversion fuel qualification team is focused on developing and qualifying plate type uranium-molybdenum (U-Mo) monolithic and uranium silicide (U3Si4) fuel. Several irradiations of mini-plates and full-size plate irradiations will demonstrate the fuel’s behavior. Along with qualifying fuel, larger tests are planned to support licensing fuel elements. These tests are referred to as DDEs. As a result of a scoping study performed in 2018, a decision was made to irradiate two DDEs in ATR at the Idaho National Laboratory (INL) and two other DDEs in BR2 in Belgium. The objective is to demonstrate successful irradiation of each DDE to support the licensing submittal of that specific reactor. Neutronics analysis is being performed to match the power density distribution of the represented elements. Test train hardware design are being done to ensure all the target conditions are being met for the two DDEs to be irradiated.
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Technical Program Plan for INL Advanced Reactor Technologies Advanced Gas Reactor Fuel Development and Qualification Program
High-temperature gas cooled reactors (HTGRs) are graphite moderated nuclear reactors cooled with helium. Their high outlet temperatures and thermal energy conversion efficiency enable efficient and cost effective integration with non electricity generating applications. These applications include process heat and hydrogen production for petrochemical and other industrial processes that require operating temperatures between 300 and 900°C. HTGRs will supplement the use of premium fossil fuels such as oil and natural gas, improve overall energy security in the United States by reducing dependence on foreign fuels, and reduce carbon dioxide (CO2)/greenhouse gas emissions. The HTGR design uses helium as a coolant, graphite as a neutron moderator, and ceramic particle fuel. Helium is chemically inert and neutronically transparent. The graphite core slows down the neutrons, retains its strength at high temperatures, provides structural stability, and acts as a substantial heat sink during transient conditions. The ceramic particle fuel is extremely robust and retains the radioactive by products of the fission reaction within the coated particle under normal and off normal conditions.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Technical Program Plan for INL Advanced Reactor Technologies Advanced Gas Reactor Fuel Development and Qualification Program
High-temperature gas cooled reactors (HTGRs) are graphite moderated nuclear reactors cooled with helium. Their high outlet temperatures and thermal energy conversion efficiency enable efficient and cost effective integration with non electricity generating applications. These applications include process heat and hydrogen production for petrochemical and other industrial processes that require operating temperatures between 300 and 900°C. HTGRs will supplement the use of premium fossil fuels such as oil and natural gas, improve overall energy security in the United States by reducing dependence on foreign fuels, and reduce carbon dioxide (CO2)/greenhouse gas emissions. The HTGR design uses helium as a coolant, graphite as a neutron moderator, and ceramic particle fuel. Helium is chemically inert and neutronically transparent. The graphite core slows down the neutrons, retains its strength at high temperatures, provides structural stability, and acts as a substantial heat sink during transient conditions. The ceramic particle fuel is extremely robust and retains the radioactive by products of the fission reaction within the coated particle under normal and off normal conditions.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Technical Program Plan for INL Advanced Reactor Technologies Advanced Gas Reactor Fuel Development and Qualification Program
High-temperature gas cooled reactors (HTGRs) are graphite moderated nuclear reactors cooled with helium. Their high outlet temperatures and thermal energy conversion efficiency enable efficient and cost effective integration with non electricity generating applications. These applications include process heat and hydrogen production for petrochemical and other industrial processes that require operating temperatures between 300 and 900°C. HTGRs will supplement the use of premium fossil fuels such as oil and natural gas, improve overall energy security in the United States by reducing dependence on foreign fuels, and reduce carbon dioxide (CO2)/greenhouse gas emissions. The HTGR design uses helium as a coolant, graphite as a neutron moderator, and ceramic particle fuel. Helium is chemically inert and neutronically transparent. The graphite core slows down the neutrons, retains its strength at high temperatures, provides structural stability, and acts as a substantial heat sink during transient conditions. The ceramic particle fuel is extremely robust and retains the radioactive by products of the fission reaction within the coated particle under normal and off normal conditions.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor
In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)
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Probabilistic Risk Assessment Comparison of Light Water Reactors and Advanced Reactors
This is my internship project about the comparison of probabilistic risk assessments for light water reactors and advanced reactors.
99 - GENERAL AND MISCELLANEOUS↗
Methodology for Stress-Controlled Fatigue Test Under In-Air and Pressurized Water Reactor Coolant Water Condition and to Evaluate the Effect of Pressurized Water Reactor Water and Loading Rate on Ratcheting [Methodology for Stress-Controlled Fatigue Test Under In-Air and PWR Coolant Water Condition and to Evaluate the Effect of PWR Water and Loading Rate on Ratcheting]
Here, this work investigates the behavior of 316 stainless steel (SS) under stress-controlled low cycle fatigue loading. Several fatigue experiments are conducted under different environment such as in air at 300°C and primary loop water conditions for a pressurized water reactor (PWR). Two different loading conditions are also employed to examine the effect of stress rate on material hardening and ratcheting. During PWR water test, actuator position measurements are used to determine the strain of the specimen. Under PWR environment, 316 SS is found to ratchet to a significantly greater degree compared with in air. At slow stress rate, higher amount of cyclic hardening is observed in 316 SS, and slow stress rate increases the rate of ratcheting. Results also indicate that 316 SS exhibits asymptotic strain response at higher stress loading which can cause material to behave very differently under same stress cyclic loading.
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Summary of the Advanced Reactor Design Criteria (ARDC) Phase 1 Activities, including the development of the Final Report and the Advanced Reactor Technology Training
Provide summary of the Phase 1 activities (Develop Final Report and Conduct Advanced Reactor Technology Training) that were completed in Fiscal Year 2015.
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Evaluation of Non-electric Market Options for a Light-water Reactor in the Midwest. Light Water Reactor Sustainability program
This study has addressed the technical feasibility and economic viability of LWR hybrid system operations based on realistic market around a specific reactor site in the Upper Midwest United States within the Great Lakes Region and near an industrial center that could benefit from more direct use of nuclear energy. The diverse industrial user and the electricity grid to optimize the revenue of the nuclear plant, depending on specific day-ahead electricity-grid capacity commitments and reserve capacity agreement requirements. For this market arrangement to work, the non-grid user is sold electricity without paying grid service fees (i.e., being considered a house load). This mode of energy sharing may require approval of governing utility commissions, depending on whether the hybrid operations can affect grid supply and manufacturing, agriculture, and transportation systems in this region provides several advantages to converting a nuclear plant into a hybrid plant that produces hydrogen and other feedstock commodities, such as polyethylene or formic acid, for example. Large-scale hydrogen production can essentially service several refineries within 20-30 miles of the nuclear plant. This would results in a measurable reduction of air-pollutant emissions reduction in the region. In addition, a large, consistent supply of hydrogen would draw new industries to the area—especially other steel-reduction and ammonia plants. For example, the nation’s first synthetic-fuels plant could also be sited near this hydrogen plant to convert CO 2 collected from the several ethanol plants in the region into methanol-based chemical feedstock and biofuels. With these perspectives in mind, this study has evaluated the business case for the following hybrid plant options.
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Advanced Reactor Technologies (ART): Very High Temperature Reactor (VHTR) Research and Development (R&D) Quarterly Report
Highlights of Advanced Gas Reactor (AGR) activities during January through March 2019.
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Microstructural characterization of cold-worked 316 stainless steel flux thimble tubes irradiated up to 100 dpa in a commercial Pressurized Water Reactor
Two flux thimble tubes (FTT) made of 15% cold-worked 316 stainless steel (SS) were harvested from Ringhals Pressurized Water Reactor (PWR) Unit 2, with peak damages of 76 and 100 displacements per atom (dpa) after 29 and 34 years’ service, respectively. Specimens sectioned from parent tubes were comprehensively characterized with nominal damage levels of ~0, ~41, ~74, 76, and 100 dpa at a nominal temperature range of 285-323 °C. Both FTTs contained helium and hydrogen gases as transmutation products. The helium follows a production rate of ~9.8 appm/dpa, while environmental factors complicate hydrogen production obscuring an exact H/dpa ratio. Irradiation-induced dislocation loops, nano-cavities, solute clusters, and microsegregation were all observed. The dislocation loops and nano-cavities indicated saturation at 41 dpa. The solute clusters continued to evolve with Ni-Si clusters formed at 41 dpa, and Ni-Si-Mn-P clusters formed at 74 and 100 dpa, but neither clusters exhibited distinct diffraction patterns at any damage levels. Solute clusters were observed to frequently be co-located with dislocation loops, but fully decorated loops were rarely detected. Significant radiation-induced segregation (RIS) was observed around grain boundaries at all damage levels. The modified inverse Kirkendall (MIK) model captured the RIS behavior of major elements. Large cavities within or around an Mn-S rich region were observed for the first time. Through all the damage levels, void swelling is always below 0.05%, making significant dimensional change unlikely in core internals when used at similar conditions. Meanwhile, the role of overwhelming nanocavities, presumably helium bubbles, should be considered in other potential degradation mechanisms, including irradiation-assisted stress corrosion cracking, embrittlement, and loss of fracture toughness, which remain the concerns for extended operation of nuclear power plants.
Post Irradiation Examination, Isotope Dilution Mas↗
Full-scale 3-D finite element modeling of a two-loop pressurized water reactor for heat transfer, thermal–mechanical cyclic stress analysis, and environmental fatigue life estimation
This paper discusses a system-level finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent sequentially coupled thermal-mechanical stress analysis were performed for typical thermal-mechanical fatigue cycles. The in-air fatigue lives of example components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in US-NRC report: NUREG-6909.
Full-scale 3-D finite element model, Pressurized w↗
The thermal decomposition of the benzyl radical in a heated micro-reactor. I. Experimental findings
Not Available
Buckingham, Grant T.↗
Research, Development and Demonstration (RD&D) Needs for Light Water Reactor (LWR) Technologies A Report to the Reactor Technology Subcommittee of the Nuclear Energy Advisory Committee (NEAC) Office of Nuclear Energy U.S. Department of Energy
The LWR RD&D Working Group developed a detailed list of RD&D suggestions and recommendations, which are provided in Appendix D. The Working Group then undertook a systematic ranking process, described in Appendix E. The results of the ranking process are not meant to be a strict set of priorities, but rather should provide insight into how the items generally ranked within the Working Group. Future discussions and investigation into these items could provide information that would support a change in these priorities or in their emphasis. The results of this prioritization are provided below. Note that in general, many RD&D ideas are applicable to both new Advanced Light Water Reactor (ALWR) plants and currently operating plants.
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Gamma-ray Analysis of the Pressurized-Water-Reactor and Boiling-Water-Reactor Assemblies
Conclusions: Passive gamma analysis can be used to determine BU, CT, and IE of spent fuel assembly. Assemblies with small CT give best results.; Without any operator declaration, a 10 minutes measurement time of a PWR assembly can give the results for BU, CT, and IE with about 3% accuracy.; BWR analysis is more complicated and less effective than PWR and the results are somewhat less accurate.; A set of 4 measurements at 4 corners of an assembly can average out the variations of the assembly positions in the reactor during the burn cycles and can result in a better accuracy.
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Nuclear reactor target assemblies, nuclear reactor configurations, and methods for producing isotopes, modifying materials within target material, and/or characterizing material within a target material
Target assemblies are provided that can include a uranium-comprising annulus. The assemblies can include target material consisting essentially of non-uranium material within the volume of the annulus. Reactors are disclosed that can include one or more discrete zones configured to receive target material. At least one uranium-comprising annulus can be within one or more of the zones. Methods for producing isotopes within target material are also disclosed, with the methods including providing neutrons to target material within a uranium-comprising annulus. Methods for modifying materials within target material are disclosed as well as are methods for characterizing material within a target material.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗