High Performance Computing User Group Meeting - June 12, 2024
INL HPC team provided updates related to the INL HPC program and HPC systems for HPC User Group Quarterly Review - June 12, 2024
99 GENERAL AND MISCELLANEOUS↗
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High Performance Computing User Group Meeting - June 12, 2024
INL HPC team provided updates related to the INL HPC program and HPC systems for HPC User Group Quarterly Review - June 12, 2024
99 GENERAL AND MISCELLANEOUS↗
Refining Absorber Shroud Geometry to Maximize Power Output and Reduce Power Peaking in ATF Test Train
In the wake of the Fukushima Daiichi nuclear power plant accident in 2011, the accident tolerant fuels (ATF) program was initiated to enhance the safety of light-water reactor fuels, placing significant emphasis on cladding. A crucial step for the broad implementation of ATF in commercial reactors involves irradiation testing of the fuel designs. ATF-2D, the latest experiment in the ATF series, is slated to undergo irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). ATF-2D is a joint effort of the INL with industry partners General Electric Global Research; Framatome; General Atomics; the Japan Atomic Energy Agency; Hitachi-GE Nuclear Energy, Ltd; Global Nuclear Fuel-Japan Co., Ltd; Nippon Nuclear Fuel Development Co., Ltd; and Mitsubishi Heavy Industries, Ltd. The ATF-2D test train design consists of four tiers, each housing six rodlets. The device will be inserted in Loop 2A within the central flux trap of the ATR, and is anticipated to undergo irradiation throughout three 60-day cycles. Typical pressurized water reactor conditions will be emulated during the irradiation. The objective of the work presented here is to dimension neutron-absorbing hafnium (Hf) components surrounding the fuel rodlets, such that the axial power profile is flattened, while simultaneously ensuring that the total fission power output of the entire test train remains below 200 kW.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Lead Service Line Inventory at the Materials and Fuels Complex
In 1986, Congress amended the Safe Drinking Water Act prohibiting the use of lead in pipes for providing water. This was due to research on lead poisoning showing lead was extremely detrimental to the development of kids and cognitive abilities in adults. Now, because of the amendment of 1986, the EPA 2021 Lead and Copper Rule Revision came out since then to promote progress in the inventory and elimination of lead pipes. The Idaho Department of Environmental Quality(IDEQ) adopted this rule for the state which means every water system in Idaho fall under the new regulation. By October 16, 2024, the INL must report to the IDEQ, the pipe materials in a Lead Service Line Inventory(LSLI) in hopes of finding lead. If lead is found, a Lead Service Line Replacement Plan is required. This poster only specifically talks about the results for the largest site out on the desert, Materials and Fuels Complex(MFC), as the INL also consists of ATR, INTEC, SMC, and CFA.
54 ENVIRONMENTAL SCIENCES↗
Radiological Mapping of Nuclear Facilities
Identifying radiological abnormalities is critical during emergencies at nuclear facilities. This facilitates a necessity for the Emergency Management (EM) to map baseline radiation levels at the Idaho National Laboratory (INL). The INL uses scintillation detectors to measure radioactivity and gamma spectrums; but, because they are heavy and costly, these detectors can be impractical for mapping radiological areas. This presentation will establish initial radiological heat maps for emergencies and test the Radiacode-103G — a handheld scintillator with GPS and Bluetooth capability with a Gadolinium Aluminum Gallium Garnet (GAGG) crystal.
61 RADIATION PROTECTION AND DOSIMETRY↗
NASA Dragonfly Mission MMRTG
The Radioisotope Power Systems (RPS) group at the Idaho National Laboratory (INL) is supporting NASA’s New Frontier class mission, Dragonfly, to explore the Saturn moon Titan. Titan has been observed by both Voyager spacecrafts in 1979 and 1980 and during over 100 close flyby encounters by the Cassini orbiter and subsequent landing of the Huygens probe which landed on Titan in 2005. The probe lasted about 2.5 hours without an RPS due to the thick atmosphere and cold climate but gave scientists valuable insight into the geology of Titan. The surface is composed of vast sand dunes and liquid methane with an atmosphere composed primarily of nitrogen that is 60% more dense as the Earth’s with 1/7th of Earth’s gravity. These conditions allow for exploration using a dual-quadcopter. Due to the extreme temperatures (-290°F or -180°C), changes to the Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) are required. This generator, designated flight unit 4 or MMRTG-F4, is being built by a team lead by INL and includes the prime contractor Aerojet Rocketdyne and their supplier Teledyne Energy System. This poster outlines the mission, the unique challenges and updates the build progress.
42 ENGINEERING↗
Multi-scale simulation of high burnup UO2 nuclear fuel during loss-of-coolant accident conditions
To improve the economics of commercial nuclear energy generation, U.S. utilities are currently seeking licensing approval to operate UO2 fuel to higher burnups. One significant safety issue that must be addressed to obtain approval is the potential for fine fragmentation/pulverization of the fuel during a loss-of-coolant accident (LOCA). It has been hypothesized that pulverization is caused by the rapid increase of pressure in fission gas bubbles in the high burnup region of the fuel during a LOCA. To better understand this phenomenon, a novel phase-field model of the fission gas bubble microstructure in UO2 has been developed and implemented in Idaho National Laboratory (INL)'s Marmot application for phase-field simulation of nuclear materials. Simulations of the bubble response to steady-state and transient conditions were conducted, and the results were used to inform a mechanistic model of pulverization in BISON, INL’s fuel performance simulation code.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Instrumentation for the In-Core Real-Time Mechanical Testing of Structural Materials (INCREASE) Project
Idaho National Laboratory (INL), in collaboration with the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), the French Atomic and Alternative Energies Commission (CEA), the Joint Research Center (JRC), the Nuclear Research and Consultancy Group (NRG), and the Research Center Rez (CVR), started a Joint Experimental Program (JEEP) project that operates within the Nuclear Energy Agency’s Framework for Irradiation Experiments (FIDES II) program in order to develop capabilities for the in-core real-time mechanical testing of structural materials. This effort will focus on designing a shared capsule capable of housing a variety of in-core mechanical testing instrumentation allowing enhanced experiments for the material science community. The outcome of the project would be high-priority, stress relaxation data for stainless-steel-based materials provided by EPRI and CEA. Stress relaxation is a major phenomenon that contributes to material degradation in nuclear reactor components. Currently, nuclear material stress relaxation is assessed both before and after irradiation, using complex and costly post-irradiation examination (PIE) activities. In-situ data would support the development of precision modeling and simulation of this degradation phenomena and would provide validation and benchmarking for existing models using the PIE data. As part of the U.S. Department of Energy Advanced Sensor and Instrumentation (ASI) program, INL has fabricated and tested out-of-core mechanical test instrumentation. This instrumentation was designed for easy adaptation to the irradiation capsule proposed under this JEEP, and can be deployed to measure real-time stress relaxation under pressurized-water reactor (PWR) conditions. This initial effort will partially serve to replace the testing capabilities lost because of shutting down the Halden Boiling Water Reactor (HBWR). The project would provide these capabilities to the international community via a shared capsule design that is easily adaptable to additional material test reactors. The design features will incorporate expansion to PWR and non-light-water reactor (LWR) environments that will be developed in future work. The capsule and instrumentation will be demonstrated in the Massachusetts Institute of Technology Reactor (MITR) for phase I and the Petten High Flux Reactor (HFR) for phase II irradiations to deliver real-time stress relaxation data on high priority stainless steel structural materials.
36 MATERIALS SCIENCE↗
Analysis of Lobe Power Calculator and Indication System with Physics and Cycle Based Models
The Advanced Test Reactor (ATR) at INL measures reactor power through two methods, thermal and Nitrogen-16 (N-16) activity. Water power calculator (WPC) is a thermal power system that measures flow and temperature to determine the thermal quadrant powers. The N-16 system utilizes a beta chamber detector that outputs reactivity levels to calculate lobe power through an algorithm called lobe power calculation and indication system (LPCIS). The LPCIS utilizes the N-16 system and the WPC system to determine reactor core power levels. The WPC provides accurate calculations of quadrant and total reactor thermal power. With the use of WPC measurements, thermal-to-N-16 (T2N) power ratios are produced to determine if the two indication systems agree on core power. Relative magnitude equations are used to utilize N-16 coefficients and multipliers to improve the indications of the LPCIS. This is crucial for maintaining safety limits. Currently, the LPCIS system uses linear equations and matrices to calculate lobe power through multipliers and coefficients. Advancements in technology and system upgrades have assisted system engineers at INL with the objective to reach a more dynamic system. In return the system demonstrates an increase in the accuracy of power reading while maintaining safety margins. The new proposed coefficient and multiple method implements a cycle specific, and a physics-based model intended for changing coefficients during operation. This calibration experiment focuses on power splitting, outer shim and neck shim, as well as fuel burning into the RDAS weighting factors. Results show that the Physics learning method yields a smaller error margin inside of the desired power range for the data set 166-A. This proved true for both constrained and unconstrained testing. This is most likely due to the physics data fitting approach that resulted in favorable coefficients and multipliers for 166-A. Continuing to improve the physics-based model will help improve the power accuracy of the LPCIS system.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Performance Investigation of High-temperature Latent Heat Storage for Integration with Advanced Nuclear Reactors
Integrating thermal energy storage (TES) with advanced nuclear reactors enhances the flexible use of nuclear energy, facilitating its expansion beyond conventional electricity generation. High-temperature TES, capable of storing heat above 400 degree C, emerges as a vital carbon-free energy solution for decarbonizing industrial sector, especially when combined with low-emission energy sources like advanced nuclear reactors. Idaho National Laboratory (INL) has recently developed a novel design for high-temperature latent heat storage system, called Heat pipe Integrated Thermal Battery (HITB), and efforts are underway to experimentally demonstrate the concept. HITB employs liquid-metal heat pipes to establish thermal linkage between TES and nuclear systems without direct fluid exchange, minimizing the potential risk of the integrated systems such as cross-contamination. HITB is designed to achieve high charging and discharging efficiency as well as high energy storage density by employing metal alloys as Phase Change Material (PCM). Initial proof-of-concept experiments are being undertaken using an aluminum alloy (Al59%-Mg35%-Zn6%). This paper discusses current progress of the HITB project at INL, seeking to develop high-temperature TES for versatile integration with advanced nuclear reactors, and shares insights from the performance evaluations via numerical modeling and analysis.
25 ENERGY STORAGE↗
Non-LWR Regulatory Framework Modernization
This report provides an end-of-year summary that reflects the progress and status of Idaho National Laboratory’s (INL) activities concerning the development of advanced reactor (AR) regulatory framework and its implementation in the United States (U.S.). The report also provides recommendations for work to be performed in Fiscal Year 2025 (FY-25) and beyond. This work was completed in Fiscal Year 2024 (FY-24) and was supported by the U.S. Department of Energy (DOE) Regulatory Development sub-program. These activities are managed by INL on behalf of DOE.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Physical Properties of Hilti HIT-RE 500 V3 After Irradiation: Slant Shear, Tensile, Flexural, and Compressive
There are various applications in which epoxy adhesives may be used within hot-cell environments. These epoxies currently see limited use within hot cells since it is known that organics see degradation after small doses even as low as 50 kilogray (kGy). These losses in physical property strengths limit the use of these epoxies to non-structural applications. In modern construction, however, the use of epoxies to bond structural components such as rebar or threaded anchors is becoming increasingly prevalent. To employ these types of materials the Idaho National Laboratory (INL) has begun investigating the specific degradation properties of epoxies for use in structural applications in environments where radiation is present. Hilti HIT-RE 500 V3 is used extensively in new construction and renovation projects across the world. The INL has conducted various tests on this epoxy for low dose environments. These tests include tensile, flexural, compression, and slant shear tests following the American Society for Testing Material International (ASTM) standards D638, D790, D695, and C882/882M respectively. Two sample sets for each test were created where the baseline specimens were not irradiated, and another set was irradiated to an accumulated dose of approximately 50-60 kGy. All the samples were destructively tested and analyzed focusing on the change in physical properties, examples of these tests are shown in Figures 1-4. Tensile testing results indicate a 0.9% loss in maximum load capacity, a 0.5% loss in tensile stress at maximum load, a 1.4% loss in tensile strain at break, and a 15% decrease in modulus of elasticity. Flexural testing results display an increase in all the following: 12% in strain at maximum load, 10% in maximum load capacity, 5% in stress at maximum load, and 3% in young’s modulus. The compressive physical properties after irradiation display a 1% increase in maximum load, a 3% decrease in stress at yield, and a 5% decrease in modulus of elasticity. The maximum compressive stress saw no measurable change after irradiation. The slant shear specimens after irradiation displayed a 2% increase in maximum load capacity and maximum compressive stress. Though a structural analysis would have to be completed as is normal for building design the increases and decreases in the measured physical properties indicate that this material may be used for structural applications in low dose hot-cell environments without adhesive failure.
36 MATERIALS SCIENCE↗
Report on the Incorporation of Post-Irradiation Examination Results of Yttrium Hydride for Advanced Moderator Handbook
Department of Energy’s (DOE’s) Microreactor program (MRP) aims to provide the fundamental data to enable the development of microreactors. As such, material property data of critical materials for microreactor technologies are researched. Because substoichiometric yttrium dihydride (YHx, where x<2) is considered as a potential solid neutron moderator, its material property data has been combined in the Advanced Moderator Material Handbook which includes thermodynamic and thermophysical properties of YHx with the exception of irradiated material’s properties due to limited PIE. To fill the knowledge gap for the irradiated YHx, specimens and irradiation capsules were prepared at Los Alamos National Laboratory (LANL). Specimens were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory’s (INL’s). post-irradiation examination (PIE) was performed at INL’s Materials and Fuels Complex (MFC). This report compiles the PIE results of irradiated YHx specimens through fiscal years 2022 and 2023 (FY22-23) . The PIE data will directly be incorporated into the newer version of the Advanced Moderator Material Handbook. The main takeaways include that (i) the geometrical stability and mechanical integrity of YHx was intact with couple exceptions after high-temperature irradiations (600-800°C), (ii) hydrogen content variation due to manufacturing or irradiation in YHx caused visible surface discoloration, that is also related to the microstructural changes, (iii) qualitative comparisons of PIE methods implied that H retention was significantly higher at 600°C as compared to 800°C, (iv) thermal properties included signatures correlated with the H loss or re-gain, (v) importance of manufacturing readiness and initial as-manufactured specimens history was emphasized, (vi) the needs of targeted irradiations focusing on temperature and time parameters and very targeted PIE were specified.
08 HYDROGEN↗
FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS
High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Dynamic Modeling and Characterization of Nuclear-grade Graphite
Idaho National Labs serves as the spearhead for many innovative energy solutions to the world's energy crisis. One such solution is the INL's Microreactor which is designed to deploy to extreme/remote environments where other sources of power are either unavailable or unreliable. In order to best design these energy solutions for their operational environments, it is crucial to understand how the design, components, and materials will respond to the environmental conditions. One key material in these innovative designs is a nuclear-grade graphite known as PCEA. This study examines the behavior of PCEA graphite under dynamic loading, similar to that which may occur in extreme environments. The objective is to characterize the dynamic behavior and produce an accurate, reliable constitutive model suitable for use in simulation tools such as INL's MOOSE. Graphite specimens were tested using a Split Hopkinson Pressure Bar (SHPB) to administer the dynamic compressive load. The SHPB was charged at various pressures to produce a range of strain rates on the material in compression. Data was acquired via strain gauges on the SHPB setup, from which stress, strain, and time data were collected. Analysis revealed the stress-strain behavior of the material as well as insights into the material behavior's relationship to strain rate. Further work must continue to characterize the various other dynamic behaviors of the material which will combine to create a substantially trustworthy constitutive model for this grade of nuclear-grade graphite. Ultimately, this will allow for realistic simulation of the material in reactor designs, allowing for prediction of design weaknesses and leading to improved designs for increased resilience, security, and reliability.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Can We Use Machine Learning to Control Nuclear Power Plants?
Slides for the INL AI/ML Symposium 11.0 focused on AI/ML in instrumentation, control, and automation. This presentation was on how we can use AI/ML for control of nuclear power plants. A recording of the presentation and other INL AI/ML symposiums can be found at https://inl.gov/artificial-intelligence/.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Technology Development of a High-Capacity High-Assay Low Enriched Uranium Transportation Concept
This paper discusses the technology development (TD) efforts that led to the development of a High-Capacity High-Assay Low Enriched Uranium Transportation (HALEU) transportation concept. In 2018, the Department of Energy (DOE) Office of Nuclear Technology and Research Development tasked Idaho National Laboratory (INL) to investigate strategies to transport large quantities of HALEU. To complete this task, INL collaborated with Pacific Northwest National Laboratory and Oak Ridge National Laboratory. The project was completed in 2020, and one of the project outcomes was a transportation concept that consisting of five individual Type B packages transported on a single legal-weight truck (LWT). The total payload capacity of this concept is 1,881 kg (4,149 lb) of HALEU in the form of uranium dioxide (UO2) powder. The concept utilizes an existing Type B packaging design carrying a novel fuel basket design with an incorporated flux trap. The basket can be loaded with 18 individual fuel canisters. The research collaboration investigated the U.S. certification potential of this concept. This part of the project included evaluations of criticality safety, radiological safety, thermal safety, structural integrity, and confinement under hypothetical accident scenarios of transport. The results of these evaluations demonstrated a promising potential for U.S. certification of this concept. Eventually, the described efforts led to the pursuance and issuance of a U.S. patent, thus, protecting the associated intellectual property (IP). Current short-term goals include making this IP available to private industry partners for licensing, directly supporting DOE’s objectives of accelerating commercialization of national laboratory-generated IP. If additional funding becomes available, long-term research goals could include exploring the feasibility of transporting other uranium chemical forms (e.g., UF4) with this concept, or refining operational procedures to load or unload the packagings.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
FUEL PERFORMANCE SIMULATION OF HIGH BURNUP FUELS IN PLANNED INTEGRAL DESIGN BASIS ACCIDENT EXPERIMENTS
High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗