AMPX and SAMMY Status Report [Slides]
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73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
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AMPX and SAMMY Status Report [Slides]
Abstract not provided.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Hall B Status Report
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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.
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 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.
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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↗
Qualification of Commercially Available Silicon Carbide for Passive Thermometry in Reactor Experiments – FY23 Status Report
Silicon carbide (SiC) is often used as a passive temperature indicator for uninstrumented in-core nuclear experiments. For the past several decades, Oak Ridge National Laboratory has relied primarily on SiC thermometry manufactured by Dow Chemical Company, formerly Rohm and Haas because of the material’s high density and reduced grain boundary elements. Although Dow SiC has performed well in this application, this material is no longer commercially available, and a new supplier is needed for future experiments. To determine a suitable replacement, a study was initiated on five types of SiC from four commercial vendors. Several critical properties, including density, electrical resistivity, chemical purity, grain structure, crystal structure, and strength were measured in samples from each material. Additionally, thermometry specimens were manufactured for irradiation in the High Flux Isotope Reactor at nominal temperatures of 300°C, 600°C, and 900°C, which will be measured using continuous dilatometry at a later date. This report provides an update of commercially available SiC material characterization and the irradiation capsule design.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Status Report on Regulatory Criteria Applicable to the Use of Artificial Intelligence (AI) and Machine Learning (ML)
Although the interest in the use of artificial intelligence (AI) and machine learning (ML) in nuclear energy is increasing rapidly, at present their implementation is limited. This rapid increase in interest is not surprising considering that implementing AI and ML technology would allow for continuous monitoring, facilitate the implementation of predictive maintenance with optimized staffing plans, enable automation and autonomy opportunities that could drastically reduce fixed operation and maintenance costs, and provide training for operations and maintenance. Other industries are using AI for construction, and in the nuclear arena AI could provide great benefit in decommissioning activities. The ability of AI and ML to operate in real time vastly increases their potential impact. Before AI can be used in design, operations, or as a regulatory tool, the specifics on the regulations applicable to the use of AI for nuclear power applications need to be established. The difficulty is that the specific use cases will dictate the applicability of regulations. For example, even within the application domain associated with operations, the regulations might vary if the AI is used to create a virtual reference for plant operations or is used for training, optimization of maintenance intervals, prioritization of maintenance activities, etc. Different still is if the AI is to be used for design or setting technical specifications, which will introduce additional requirements. US Nuclear Regulatory Commission (NRC) licensing reviews are based on an applicant’s design meeting its performance assessment based on (1) safety goals and objectives, (2) deterministic and/or probabilistic analysis of accident scenarios, and (3) quantitative assessment of design alternatives against the safety goals and objectives using accepted engineering tools, methodologies, and performance criteria. The current regulatory framework does not explicitly address AI or autonomous control. However, as implementing AI technology will require the use of a digital platform, it must meet the requirements of an instrumentation and control (I&C) system. The regulatory requirements for AI, which will be incorporated into the I&C system, will be very dependent on how it is used (i.e., its functionality, safety classification, etc.). The licensing process is primarily risk-based with the identification of components and systems as nonsafety, important to safety, or safety related. A risk-informed approach allows further gradation of components and systems based on risk metrics such as core damage frequency or large early release fractions. Thus, the use cases and the risk categorization of impacted systems and components will determine the regulatory requirements. Regardless of how AI is used it presents new opportunities for risk-informing operating, maintenance, and regulatory decisions. Trustworthiness, transparency, and the ability to validate and verify the results will be paramount in showing that the systems and plant still meet their performance requirements. This report describes the results of research to identify regulatory implications of AI technologies and their uses. Specifically, this report reviews current regulatory guidance relevant to the application of AI for design (including design changes or new designs including advanced reactors), construction, operations, training, maintenance, research, testing, and as a regulatory tool. AI can be automated at different levels from purely informative purposes to autonomous controls. The focus of this review included determination of constraints on the application of AI technology, identification of any regulatory gaps or uncertainties, and clarification of anticipated technical basis information likely to be important for regulatory acceptance of these technologies. Currently, any use of AI at nuclear power plants is focused on nonsafety-related applications. The NRC and other regulatory bodies are evaluating providing guidance to address gaps rather than create new regulations to address the use of AI and ML. This approach seems to be the best to encourage AI development without adding regulatory uncertainty.
97 MATHEMATICS AND COMPUTING↗
2023 Status Report - Fiberboard Properties and Degradation Rates for Storage of 9975 Shipping Package in KAC
Thermal, mechanical, and physical properties have been measured on fiberboard samples aging for up to 18 years at accelerated conditions. The aging environments include elevated temperature up to 250 ºF (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.
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FY2023 Status Report: Model 9975 O-Ring Fixture Long-Term Leak Performance
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.
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Inspection Status Report of Site A/Plot M Report for 2023
The Site A/Plot M Decommissioned Reactor Site was inspected on June 6, 2023. Site A/Plot M is in the Palos Area Preserves, operated by the Forest Preserve District of Cook County. The site was found to be in good condition with negligible erosional concerns across the grass covered mound at Plot M. Landscaping timbers installed along the footpath of Plot M have been effective in reducing erosion down the eastern slope, thus, no additional timber steps are needed at this time. The bike trail, installed by Cook County Forest Preserve, located to the south and east of Plot M has helped to significantly reduce traffic at the site. The Site A and Plot M monuments were both replaced on May 1, 2021, and remain in good condition, despite a small amount of graffiti observed on both monuments at the time of the inspection. The 19 groundwater monitoring wells at Site A/Plot M were found to be secure and in good condition.
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Status Report of Joint RPI/ORNL NCSP Task for Thermal Neutron Total Cross Section Measurements
A new cold neutron moderator system was developed and constructed at the Rensselaer Polytechnic Institute (RPI) Gaerttner Linear Accelerator. This system was then used to measure the neutron total cross section in the thermal energy range for polyethylene, polystyrene, Lucite, and yttrium hydride. In tandem with these measurements, a fitting procedure was developed at Oak Ridge National Laboratory to vary computationally simulated phonon density of states until a fit with differential and integral data was achieved. This model, combined with the new RPI total cross section data, was used to generate thermal scattering law files for the materials measured at RPI.
36 MATERIALS SCIENCE↗
FY23 Status Report: SNF Interim Storage Canister Corrosion and Surface Environment Investigations
Work evaluating spent nuclear fuel (SNF) dry storage canister surface environments and canister corrosion progressed significantly in FY23, with the goal of developing a scientific understanding of the processes controlling initiation and growth of stress corrosion cracking (SCC) cracks in stainless steel canisters in relevant storage environments. The results of the work performed at Sandia National Laboratories (SNL) will guide future work and will contribute to the development of better tools for predicting potential canister penetration by SCC.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗