AMPX and SAMMY Status Report [Slides]
Abstract not provided.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
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Status Report 2: Advanced Nuclear Reactors Utilized for Synthetic Fuel Creation
Synthetic fuels (synfuels) are hydrocarbon fuels that source energy from electricity. Synfuels have the potential to significantly reduce greenhouse gas emissions throughout the transportation sector. To achieve this substantial reduction in greenhouse gas emissions, the electricity must be sourced from zero- or near-zero-carbon fuel sources such as solar, wind, hydroelectricity, and nuclear power. Synfuels are produced from a combination of carbon and hydrogen sources. Hydrogen can be sourced from water electrolysis with near-zero-carbon electricity and heat (e.g., nuclear), while carbon dioxide can be sourced from ethanol and ammonia plants. In the hydrocarbon fuel synthesis process, hydrogen and carbon dioxide can be reacted to produce carbon monoxide and water, via the so-called reverse watergas shift reaction. Carbon monoxide can then react with additional hydrogen to form hydrocarbons, with carbon chains ranging from C1–C30 in the reaction known as the Fischer-Tropsch (F-T) synthesis reaction. The synthesized hydrocarbon molecules can then be hydro-processed with additional hydrogen and distilled into different carbon chain lengths so as to be compatible with existing conventional gasoline, jet, and diesel fuels. Carbon-free synfuel production comes with a “green premium” over the manufacture of identical products via conventional fossil fuels. Reports from Argonne National Laboratory (ANL) reveal that hydrogen costs dominate the cost of carbon-free synfuel production. This suggests that for the cost of green synfuel to approach that of conventional petroleum fuel, the cost of hydrogen must be approximately $\$1$/kg. Of the primary low-carbon energy sources, only nuclear carries the potential to produce hydrogen at below $\$2$/kg. (Still a bit above the lofty $\$1$/kg goal, but perhaps manageable). To further identify the potential for creating low-cost synfuels capable of competing with legacy technologies, the Department of Energy Office of Nuclear Energy has funded a multi-program, multi-lab effort among ANL, Idaho National Laboratory (INL), the Integrated Energy Systems (IES) program, and the Light Water Reactor Sustainability (LWRS) program. This collaboration effort will determine the possibility of using current and next generation nuclear reactors to create low-cost carbon-free synfuels for sale in the U.S. energy and commodities market.
10 SYNTHETIC FUELS↗
Model 9975 O-Ring Fixture Long-Term Leak Performance (FY2022 Status Report)
Leak testing experiments to monitor the aging performance of Viton® GLT and GLT-S O-rings used in the model 9975 shipping package has been ongoing since 2004 at Savannah River National Laboratory. Seventy tests using mock-up 9975 primary containment vessels (PCVs) with GLT O-rings were assembled and heated to temperatures ranging from 200 to 450 °F. Due to material substitution, fourteen tests with GLT-S O-rings were initiated in 2008 and heated to temperatures ranging from 200 to 400 °F. The conditioning temperatures are elevated compared to the calculated maximum O-ring temperature in a 9975 package in storage, 158 °F, to accelerate aging and enable observations of O-ring failures in a reasonable time frame. The fixtures are leak tested periodically and all GLT O-ring fixtures aging at 350 °F or above have failed to maintain a leak-tight seal. Eight GLT O-ring fixtures aging at 300 °F have failed after 2.8 to 5.7 years at temperature while the remaining fixtures at 300 °F were retired from testing following more than five years without failure. There has been one GLT O-ring fixture which failed after 13.4 years aging at 200 °F. However, 20 other GLT O-rings aging at 200 °F have remained leak-tight for over 14 years and remain in test. No failures have yet been observed in GLT O-ring fixtures aging at 270 °F for 10.4 years. All GLT-S O-ring fixtures aging at 300 °F or above have failed their leak test. No failures have yet been observed in GLT-S O-ring fixtures aging at 250 °F for 12.5 years. This year, one GLT-S O-ring fixture failed after 12.4 years aging at 200 °F. Data from the O-ring fixtures are generally consistent with results from compression stress-relaxation testing, and provide confidence in the predictive models based on those results. However, uncertainty exists in extrapolating these elevated temperature results to the lower temperatures of interest for normal storage in K-Area Complex (KAC). Oxygen consumption testing, which includes results from temperatures near KAC normal storage temperatures, is ongoing to provide further confidence and corroborate these extrapolations. The collective data from these test efforts suggest the minimum O-ring service life at KAC normal storage conditions should be at least 34 years for GLT and GLT-S O-rings. Measurement of compression set in O-rings removed from failed fixtures, compared to that from KAC surveillance O-rings, indicate significant margin remains for O-rings still in service in 9975 packages in KAC. Aging and periodic leak testing will continue for the remaining 25 mock-up PCV fixtures.
36 MATERIALS SCIENCE↗
Fiberboard Properties and Degradation Rates for Storage of 9975 Shipping Package in KAC (2022 Status Report)
Thermal, mechanical, and physical properties have been measured on fiberboard samples following accelerated aging for up to 17 years. The aging environments include elevated temperature up to 250 °F (the maximum allowed service temperature for fiberboard in 9975 packages) and elevated humidity. Accelerated aging results have been analyzed and used to build aging models. Correlations relating several properties (thermal conductivity, energy absorption, weight, dimensions, and density) to their rate of change in potential storage environments have been developed. Combined with an estimate of the actual conditions the fiberboard experiences in K-Area Complex (KAC), these models allow development of service life predictions. KAC completed calculations that supported extending the service life of 9975 packages with 3013 containers in storage from 20 years to 40 years. These calculations addressed the potential for degraded fiberboard properties following 40 years in storage using degradation models developed previously. Due to differences in containers, 9975 packages with non-3013 containers are only approved for 20 years of storage. The updated degradation models developed in this report have been compared to the assumptions in those calculations, and the current results for fiberboard dimensions, density, radial thermal conductivity, and energy absorption remain consistent with those calculations. The axial thermal conductivity degradation rate slightly exceeds calculations estimate of 0.5% per year between 130 and 160 °F. Additional data will continue to be collected to permit future refinements to the models and assumptions. Some of the predicted degradation rates presented in this report are extreme. However, these relate to environments that do not exist within KAC or would be postulated only as upset conditions that would not likely persist for an extended period. For a typical package stored in KAC with approximately 10-12 watts internal heat load or less, and ambient temperatures below 90 °F, the fiberboard assemblies experience storage conditions less severe than any of the aging environments. Fiberboard in conforming packages with lower internal heat loads should experience little or no degradation and is expected to provide a service life beyond the currently approved 40 year storage period. Packages with higher internal heat loads may not continue to perform their required safety functions beyond 40 years. Ultimately, service life will be determined by the cumulative effect of degradation from all the conditions these packages might encounter. The results and model predictions presented in this report are applicable to 9975 packages with cane or softwood fiberboard overpack assemblies. These degradation models do not address the effects of non-conforming conditions such as the presence of excess moisture and mold or beetle infestations.
36 MATERIALS SCIENCE↗
PELICAN Status Report for Year 2021
The Pressure Drop Experimental Loop for Investigations of Core Assemblies in Advanced Nuclear Reactors, PELICAN, is a novel experimental facility designed and constructed to recreate the prototypic full scale hydraulic flow conditions for the Versatile Test Reactor (VTR) currently under development by the U.S. Department of Energy. To support VTR design efforts, PELICAN is used to replicate VTR’s primary heat transport system (PHTS) in the core assembly region. The measurement objectives for data generated from this test facility are to mimic the prototypic VTR assembly and to meet the validation needs for code calculations and simulations of the reference design for the VTR core. These objectives focus on the need to validate pressure drop results across the various segments of the fuel assembly, including the axial reflectors, fuel, and transition regions, as they relate directly to the pumping power and safety behavior of the reactor. In the following sections, the major milestones for the experimental work performed on PELICAN over the calendar year (CY) 2021 are presented. In the next section, we discuss the assembly and shakedown testing of the facility. In Section 3, an analysis of the errors in the major measurements is performed. In Section 4 and 5 we present the results from the first set of test articles placed inside the loop. The design of the remaining test articles as part of the full assembly are shown in Section 6 and the conclusion and path forward for the following year, CY22 is given in the final section.
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FY22 Status Report on the ART-GCR CMVB and CNWG International Collaborations
This work presents the numerical model for the high temperature test reactor (HTTR) loss of forced cooling (LOFC) experiment with the INL codes Griffin, BISON, and RELAP-7, based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. Promising results were obtained, with the overall behavior of the reactor successfully captured. Changes in the heat transfer coupling, as compared to the fiscal year (FY)-21 model, enabled drastic improvement of the steady-state solution, both in terms of computational time and global energy discrepancies. The former was reduced by a factor of roughly 60, whereas the latter decreased from 11% to less than 2%. Furthermore, the discrepancy in the steady-state multiplication factor was improved from +2,300 and +2,900 pcm (for the 30 and 9 MW cases) to -700 and +1,200 pcm, respectively, and now falls well within the large measurement uncertainties stemming from graphite impurities. Validation of the Monte Carlo model used to generate cross sections was also performed against available measurements. Though significant, the discrepancies remain acceptable overall in light of the large uncertainty. Numerous improvements are still needed to better compare with the experiment involving the 9 MW case and to instill greater confidence in the model’s ability to accurately predict the 30 MW behavior. Specifically, the power levels predicted by the 9 MW transient simulation following re-criticality remain low, pointing to an underestimation of the passive cooling of the core. A key aspect of future work will be to better understand the flow pattern during the LOFC event, particularly to determine if natural or forced convection is occurring inside the reactor pressure vessel (RPV). More generally, additional validation data would be immensely useful for further enhancing the numerical model and better matching the experiments. In addition, a more sophisticated thermal-hydraulics model that simulates all the channels as a single system model should be considered to take into account the rest of the primary loop. Finally, even if the results are in better agreement with the experiments, sensitivity analysis and uncertainty quantification will be necessary to evaluate the model uncertainty.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Composite Analysis for Low Level Waste Disposal in the Central Plateau of the Hanford Site (Annual Status Report, FY 2021)
In accordance with DOE M 435.1-1 requirements and as implemented by DOE/RL-2000-29, the U.S. Department of Energy (DOE), Richland Operations Office has prepared this annual summary for fiscal year (FY) 2021. Originally reported in PNNL-11800 and PNNL-11800, Addendum 1 (hereinafter collectively referred to as the Hanford Site Composite Analysis), the Hanford Site Composite Analysis was approved through issuance of a 2002 memorandum. As required by DOE/RL-2000-29, an annual evaluation of new information and data developed by a number of onsite programs was completed. The reporting period for this annual evaluation is FY 2021 (October 1, 2020 through September 30, 2021).
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
AMPX and SAMMY Status Report [Slides]
Abstract not provided.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
AMPX and SAMMY Status Report [Slides]
Abstract not provided.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Hall B Status Report
Explore the source record for details and available documents.
Achenbach, Patrick↗
Composite Analysis for Low-Level Waste Disposal in the Central Plateau of the Hanford Site (Annual Status Report FY 2022)
In accordance with DOE M 435.1-1 requirements and as implemented by DOE/RL-2000-29, the U.S. Department of Energy (DOE), Richland Operations Office has prepared this annual summary for fiscal year (FY) 2022. Originally reported in PNNL-11800 and PNNL-11800, Addendum 1 (hereinafter collectively referred to as the Hanford Site Composite Analysis), the Hanford Site Composite Analysis was approved through issuance of a 2002 memorandum.5 As required by DOE/RL-2000-29, an annual evaluation of new information and data developed by several onsite programs was completed. The reporting period for this annual evaluation is FY 2022 (October 1, 2021 through September 30, 2022).
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Performance Assessment for the Disposal of Low-Level Waste in the 200 East Area Burial Grounds (Annual Status Report FY 2022)
This annual review provides the projected dose estimates of radionuclide inventories disposed in the 200 East Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-SD-WM-TI-730). The estimates are compared with the performance objectives defined in U.S. Department of Energy requirements (DOE O 435.1 and its companion documents DOE M 435.1-1 and DOE-STD-5002-2017). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063) and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 is expected.
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Performance Assessment for the Disposal of Low-Level Waste in the 200 West Area Burial Grounds (Annual Status Report FY 2022)
This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 West Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-EP-0645). The estimates are compared with the performance objectives defined in U.S. Department of Energy requirements (DOE O 435.1 and its companion documents DOE M 435.1-1 and DOE-STD-5002-2017). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063) and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 is expected. Within the active burial grounds, low-level and mixed low-level waste currently may be disposed only in two lined trenches in the 218-W-5 Burial Ground (Trenches 31 and 34) until they are either filled or a decision is made to close these trenches. Some mixed low-level waste is also disposed at the Environmental Restoration Disposal Facility in the 200 West Area (which is covered under a separate PA). During this (fiscal year 2022) reporting period (October 1, 2021, through September 30, 2022), waste was disposed to the 200 West Area LLBGs. Continued groundwater monitoring of the 200 West Area LLBGs indicates no groundwater contamination due to LLBG waste. Current assumptions about future land use at the Hanford Site are consistent with PA analysis assumptions of a postclosure facility that will not be degraded by human activity. The LLBGs are in an area identified for waste management and containment of residual contamination (DOE/EIS-0391). The current closure plan for the LLBGs (DOE/RL-2000-70) estimates that the 200 West LLBGs will be closed in the 2050 timeframe. The Disposal Authorization Statement, other technical basis documents, and the radioactive waste management basis are of continued adequacy to meet the performance objectives of DOE O 435.1. Overall, there are no substantive changes to primary PA assumptions and no changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 and the Disposal Authorization Statement is maintained. A new PA to evaluate the long-term impacts of three disposal trenches that are currently active within the 200 East and 200 West Areas (Trench 94 in the 200 East Area and Trenches 31 and 34 in the 200 West Area) was initiated in fiscal year 2019 and completed in fiscal year 2022. Corrective actions addressing 3 key issues and 31 secondary issues identified during the review process were developed and submitted to the Low-Level Waste Disposal Facility Federal Review Group Co-Chairs for review and approval. This PA provides additional technical basis for the continued adequacy of the existing Operating Disposal Authorization Statement.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Performance Assessment for the Environmental Restoration Disposal Facility (Annual Status Report FY 2022)
DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA) (CP-60089), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and that the determination must consider the results of data collection and analysis from research, field studies, and monitoring as well as the need to update any Radioactive Waste Management Basis (RWMB) documents. Beginning in 1996, the Environmental Restoration Disposal Facility (ERDF) started accepting low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells or disposal areas and can accommodate future design expansions as needed. Currently, there are eight cells and two supercells in ERDF. Each supercell is the equivalent of two cells. During this reporting period (fiscal year 2022, extending from October 1, 2021, through September 30, 2022), approximately 8.52E+04 metric tons (9.39E+04 U.S. tons) of waste was disposed at ERDF. From ERDF inception through September 30, 2022, approximately 17.0 million metric tons (18.7 million U.S. tons) of waste has been disposed at ERDF, which equates to consumption of approximately 89.1% of the currently constructed disposal volume. According to the design of ERDF, the facility has the ability to be expanded as needed. As a condition of the DAS, disposal operations within ERDF must be in accordance with the waste acceptance criteria (ERDF-00011) that provide specific radionuclide disposal limits, waste form restrictions, and descriptions of acceptable waste packages in compliance with the requirements of DOE M 435-1.1. The ERDF waste acceptance criteria stipulate that waste destined for disposal at ERDF be controlled based on source, physical form, and contaminant concentration and activity levels. There have been no changes to the physical configuration of ERDF or to the waste forms (source, physical form, etc.). No new unreviewed disposal question screenings or evaluations have been generated in this reporting period. Therefore, there are no noted impacts to the PA, CA, DAS, or RWMB resulting from the evaluations and screenings. Sum-of-fractions analysis shows that the disposed inventory meets both the concentration and inventory threshold requirements. A sum-of-fractions value is computed for ERDF sensitive radionuclides contributing to the groundwater pathways and the air pathway inventory limits. Computed values were 5.50E-04 and 3.30E-03, respectively. The disposed waste inventory remained well under the PA imposed limits, as shown in Table 4 and Table 5 in the main text of this report. Required monitoring was satisfactorily completed during the fiscal year reporting period. Compliance with performance objectives were met as each of the reported values were well below the established limit. Overall, there are no substantive changes to primary PA assumptions or changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 and the DAS is maintained.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Performance Assessment for the Disposal of Low-Level Waste in the 200 West Area Burial Grounds (Annual Status Report FY 2021)
This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 West Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-EP-0645). The estimates are compared with the performance objectives defined in U.S. Department of Energy requirements (DOE O 435.1 and its companion documents DOE M 435.1-1 and DOE-STD-5002-2017). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063). and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 is expected.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Performance Assessment for the Environmental Restoration Disposal Facility (Annual Status Report FY 2021)
DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA) (CP-60089), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and that the determination must consider the results of data collection and analysis from research, field studies, and monitoring as well as the need to update any Radioactive Waste Management Basis (RWMB) documents. Beginning in 1996, the Environmental Disposal Facility (ERDF) started accepting low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells or disposal areas and can accommodate future design expansions as needed. Currently, there are eight cells and two supercells in ERDF. Each supercell is the equivalent of two cells. During this reporting period (fiscal year 2021, extending from October 1, 2020, through September 30, 2021), approximately 9.14E+04 metric tons (1.01E+05 U.S. tons) of waste was disposed at ERDF. From ERDF inception through September 30, 2021, approximately 16.9 million metric tons (18.9 U.S. tons) of waste has been disposed of at ERDF, which equates to consumption of approximately 88.7% of the currently constructed disposal volume. According to the design of ERDF, the facility has the ability to be expanded as needed. As a condition of the DAS, disposal operation within ERDF must be in accordance with the waste acceptance criteria (ERDF-00011) that provide specific radionuclide disposal limits, waste form restrictions, and descriptions of acceptable waste packages in compliance with the requirements of DOE M 435-1.1. The ERDF waste acceptance criteria stipulate that waste destined for disposal at ERDF be controlled based on source, physical form, and contaminant concentration and activity levels. There have been no changes to the physical configuration of ERDF or to the waste forms (source, physical form, etc.). No new unreviewed disposal question screenings or evaluations have been generated in this reporting period. Therefore, there are no noted impacts to the PA, CA, DAS, or RWMB resulting from the evaluations and screenings. Sum-of-fractions analysis shows that the disposed inventory meets both the concentration and inventory threshold requirements. A sum-of-fractions value is computed for ERDF sensitive radionuclides contributing to the all pathways and the air pathway inventory limits. Computed values were 7.63E-03 and 1.49E-03, respectively. The disposed waste inventory remained well under the PA imposed limits, as shown in Table 4 and Table 5 in the main text of this report. Required monitoring was satisfactorily completed during the fiscal year reporting period. Compliance with performance objectives were met as each of the reported values were well below the established limit. Overall, there are no substantive changes to primary PA assumptions or changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 and the DAS is maintained.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Gripper Test Assembly Status Report (FY2020)
Refueling systems for fast reactors are designed to handle fresh and used core assemblies (fuel, reflector, and shield core assemblies) within the reactor vessel in an opaque coolant environment without visual reference. These refueling machines are designed to work in a sodium (or other fast reactor coolant) and argon vapor space environment and are engineered with the rotatable plug system to allow for the movement of fresh and spent fuel into and out of the reactor core. The refueling machines are a critical component in any reactor and thus need to undergo extensive testing in a prototypic environment to ensure that they will meet all of the system functions and requirements. Argonne has developed an innovative compact refueling system design for the Advanced Fast Reactor-100 that is based upon some mechanisms used in previous reactor designs, such as the U.K.’s Prototype Fast Reactor (PFR) and some mechanisms that have not been used in sodium. This compact refueling machine supports the reduction in size of the AFR-100’s reactor vessel, and if fully developed, would support and inform the development of the in-vessel refueling machines for such commercial reactors as the GEH PRISM reactor plant, the ARC Clean Energy’s ARC-100 reactor, and the Natrium reactor, among others. This refueling system is a vital component of a fast reactor that supports reducing the cost of the reactor and increasing its reliability. During the development of the compact fuel handling machine conceptual design for the AFR-100, a lack of testing data for many mechanical components in sodium under typical in-reactor loads and conditions was discovered. The reduction in lifetime of the various mechanical components in the liquid sodium environment needs to be quantified versus the calculated component lifetimes under normal conditions in the testing while they are subjected to typical loading profiles experienced in the past. The Gripper Test Assembly discussed here includes a full-size gripper device with appropriate mechanical features that will be tested in sodium to provide this testing data. The Gripper Test Assembly is used to test various mechanical fuel handling components submerged in high temperature liquid sodium. These mechanisms are gears, bearings, gripper jaws and head, universal joints and shafts, ball screws, among others. These components will be tested under the typical sodium environmental conditions experienced during refueling operations with appropriate loading conditions that simulate the removal and insertion of core assemblies into a fast reactor grid plate structure. This Gripper Test Assembly is the second in a series of refueling system mechanisms developed for testing in sodium. The first test assembly is the Gear Test Assembly (GTA) which was used to test the performance of gears and bearings operating in sodium. Because of the successful testing conducted with the GTA, it was decided to continue with the development of the Gripper Test Assembly which uses the same gears and bearings tested in GTA. Using the data collected during operation of this gripper test assembly, lifetime reduction factors of the various mechanical components can be calculated for the material combinations selected. These lifetime reduction factors can be used in the design of future mechanical systems which operate in these environments to accurately predict component end of life. In addition, an understanding of the ability of these components and mechanisms to operate under-sodium with the chosen materials will be accomplished. Again, this Gripper Test Assembly is the follow-on test article to the Gear Test Assembly which was testing the ability of gears and bearings operating in a sodium environment. Once fabricated and qualified, it will be tested in the Mechanism Engineering Test Loop (METL) facility located in Building 308 at Argonne National Laboratory.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
UCO TRISO Minifuel FY23 NSUF-Kairos Power Post-Irradiation Examination Status Report
Irradiation of miniature tristructural isotropic (TRISO)–coated particle fuel compacts at high-power particle was performed in the Oak Ridge National Laboratory’s (ORNL’s) High Flux Isotope Reactor (HFIR) using the MiniFuel irradiation capability. Each compact comprised 20 TRISO particles with a low-enriched uranium carbide uranium oxide (UCO), natural UCO, or low-enriched UO 2 kernel within a graphitic matrix. After irradiation, the MiniFuel targets and subcapsules were disassembled to recover the irradiated fuel specimens and pursue post-irradiation examination (PIE) to inform Kairos Power on the fuel specimen performance. This report describes the PIE results collected to date, including dilatometry on the passive thermometry to confirm the irradiation temperature, fission gas release measurements, and gamma counting. This work was funded by the Nuclear Science User Facilities program.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗