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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Diffusional creep in UO 2 informed by lower length scale simulations

Using molecular dynamics, we predict information at the atomistic scale used to develop a mechanistic UO 2 creep model for use in higher length-scale fuel performance codes. The ultimate objective of the model is to not only to capture the creep rates of UO 2 but to determine the dominant mechanism in the diffusional regime, which is still debated in the literature. It is important to have a model to capture the correct mechanisms for creep in UO 2 as this can be used as the foundation when applying to other fuels, such as doped UO 2 , and when irradiation is accounted. In last years NEAMS milestone (FY22), we developed a prelimnary model, however there were issues, for example, excessively high values of uranium vacancy concentrations at the grain boundary. This year we have addressed the issues with the previous version of the model, added a new term that accounts for the nucleation of dislocations at stress raisers (e.g., triple junctions) within the microstructure and discussed where there was disagreement in the literature about the underpinning physics (uranium self-diffusion at the grain boundary).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

The Impact of the Halden Reactor Project on the Development of the Bison Fuel Performance Code

The Bison fuel performance code has been under development at Idaho National Laboratory by various programs within the U.S. Department of Energy since 2009. As part of any fuel performance code development, validation is necessary. To this end, several Halden experiments have been used. Early on, the focus was on validating the thermal mechanical behaviour, including pellet cladding mechanical interaction and fission gas release models implemented in Bison. Support for design-basis accident capabilities began in approximately 2014. Also in 2014, the development of models for accident tolerant fuel (ATF) concepts also began. This includes dopedUO2 fuel for which Halden has some experiments with temperature and fission gas release measurements. The Bison team has been fortunate to have two senior developers be secondees to Halden for a year: the late Dr. Giovanni Pastore from May 2015 to April 2016 and the now retired Dr. Richard Williamson in 2018. The paper will provide examples of Bison validation using Halden data for standard UO2 fuel under normal operation and loss-of-coolant accident condition, and ATF concepts. A tribute to the legacy that both the late Giovanni and retired Richard left on the Bison code is provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Damage progression and failure of SiC/SiC composite tubes under hard-contact radial expansion

The response of silicon carbide (SiC) fiber-reinforced SiC matrix (SiC/SiC) composite cladding to mechanical interaction with fissile fuel is a knowledge gap that must be overcome to design and assess SiC-based cladding systems for advanced nuclear applications. This study developed the relevant mechanical testing capability and identified the failure behavior and the critical microstructural features and processing defects. Sections of SiC composite tube were subjected to a modified expansion-due-to-compression (EDC) test in an X-ray computed tomography microscope: a polyurethane plug pressed surrogate Al 2 O 3 into the inner walls of the SiC/SiC composite tubes to achieve hard contact. A pure EDC test with just a polyurethane plug was also performed as a reference. Through the use of displacement fields, digital volume correlation revealed inhomogeneous deformation fields in the tubes, even for pure EDC, which was related to the inherent defects in the structure. Deep learning–aided segmentation and systematic data analysis revealed that the presence of inhomogeneous deformation applied by the hard contact was exaggerated by the presence of inner surface imperfections left behind from the matrix densification process. In conclusion, the findings provide insights into the applications, highlighting the necessity for improvements in inner surface roughness and the incorporation of localized contacts in pellet–cladding mechanical interaction computational models.

Composites↗
DOE OSTI2023

Simplifying the Advanced Test Reactor LOWE Element Test Series

The Advanced Test Reactor (ATR) is one of six remaining high performance research reactors in the United States scheduled to be converted from 93% high-enriched uranium (HEU) fuel to 19.75% low-enriched uranium (LEU) fuel as part of the Department of Energy (DOE), National Nuclear Security Administration (NNSA), Office of Material Management and Minimization (MMM) reactor conversion efforts(1). The specific LEU fuel element design for the ATR is called the Low Enriched (LOWE) element. A series of fuel qualification tests are scheduled to occur over the next decade before the ATR is fully converted. In addition to several experiments of demonstration plates, a series of Element Tests (ETs) are planned for conversion, in which full sized LOWE elements are placed in ATR driver positions. The purpose of these ETs was generally to sequentially increase the amount of LEU and core power to full power, therefore creating a representative safety profile in which LEU could operate for the duration of the ATR lifetime. The planned element tests were (2): ET-1: one LOWE element at low power for once cycle ET-2: ~8 LOWE elements at low/medium power for multiple cycles, and ET-3: >8 LOWE elements for a full lobe of elements at high power for multiple cycles. Given recent improvements in modeling fidelity, scheduling considerations, and an opportunity to combine later ETs, the LEU conversion program successfully defined the operational requirements for the “ET-ATR” test, which combines the needed information collected from ET-2 and ET-3 into a single test.

42 ENGINEERING↗
DOE OSTI2022

The Mechanical Response of High Burnup 17x17 PWR Fuel Rods Under Bending

As a part of the US Department of Energy Office of Nuclear Energy High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory has performed 4pt-bend tests on a series of 17x17 PWR fuel rods that were discharged with rod average burnups between 50 and 60 GWd/MTU. The rods include four different kinds of fuel cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Six rods of each type (combining Zirc-4 and LT Zirc-4 as one type) were tested to failure at room temperature and 200 oC. Three of each type experienced a full-length heat treatment cycle from RT to 400 oC to RT with an 8 hr soak at 400 oC to simulate the expected limiting vacuum drying condition, which potentially induces hydride reorientation and causes some annealing of irradiation damage in the cladding. Overall, the fuel rods show significant capacity to deflect in bending and exhibit non-linear deformation before failing. The calculated failure strains range from 1.5% to 3.8%. The simulated vacuum drying heat treatment had the largest effect on fuel rods with ZIRLO and M5 cladding. In both cases the heat treatment caused increases in the plastic strain at failure, suggesting some annealing of irradiation damage is occurring. The finite element modeling of the 4pt bending test assume the pellet is a collection of pellet wedge pieces. The results for the Zircaloy-4 rods suggest that the pellets are bonded to the cladding and friction controls the relative motion of the pellet wedge pieces.

Cantonwine, Paul↗
DOE OSTI2022

Thermal-hydraulic and Fuel Performance Scoping Studies of a Flowing Water Capsule in TREAT

The restart of the Transient Reactor Test (TREAT) facility has provided a much needed capability for integral transient testing of nuclear fuel. This testing is necessary to qualify new fuel concepts such as those developed under the Accident Tolerant Fuel (ATF) program, increasing the burnup limit of light water reactor (LWR) fuels, or a variety of other programs under the Advanced Fuels Campaign (AFC). The ATF campaign has been the main driver behind the development and implementation of a variety of capsules for TREAT experiments. The Separate-Effect Test Holder (SETH) was a dry capsule that enabled testing of ATF concepts under reactivity-initiated accident (RIA) heating conditions. Following SETH, the Static Environment Rodlet Transient Test Apparatus (SERTTA) was developed to enable RIA testing in a static water environment. In efforts to support the need for future Loss-of-Coolant Accident (LOCA) tests, the Transient Water Irradiation System for TREAT (TWIST) capsule has been developed that allows for water to drain from around the fuel rod and lower the pressure to simulate LOCA conditions. All these capsules that have been developed for LWR fuel testing all lack the capability for forced convective cooling which in some applications may limit their ability to test under prototypic conditions. A design and modeling effort has been started to modify the TWIST capsule by adding a flow tube and impeller that can create flowing coolant conditions for the fuel rod. RELAP5-3D and BISON models are being used to study the differences between RIA, LOCA, Anticipated Operational Occurrences (AOO), LWR power cycling, and other transient scenarios under flowing conditions capable in the flowing capsule and the current stagnant water capsules (SERTTA and TWIST). The scoping study will provide guidance on the needed capabilities for the flowing capsule and the limitations of the currently developed capsule for LWR testing in TREAT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Considerations for Improving the Technoeconomic Viability of Heat Pipe Microreactors by Employing a Functional Containment Approach

The economic cost of a microreactor is being investigated by leveraging MARVEL cost information in conjunction with some simplified design and analysis activities to present a hypothetical, yet feasible, option for investigation. This bottom-up cost estimate for a hypothetical commercial microreactor will be used to provide guidance to the industry and research communities regarding where priorities should be set to improve the viability of broad microreactor deployment. To support the design and analysis activities required to present a feasible option, some understanding of maximum accidental radionuclide releases and consequences is needed to provide bounding estimates on containment or confinement systems for radionuclide retention. To estimate fission product barrier and retention performance, a gaussian plume dose and dispersion model is employed, with isotopic inventories being propagated between barriers assuming conservative fractional release rates. Demonstrably conservative meteorological assumptions are also assumed. Standard reference release rates are assumed where safety requirements could be easily and economically employed. Results are provided in units of total effective dose equivalent as a function of distance from the release (i.e., the reactor). Initial results indicate that the hypothetical microreactor will likely need to credit the fuel cladding and at least one other barrier, or combination of barriers, to ensure public health and safety as required during a postulated accident. This approach is referred to as “functional containment,” since no single barrier will be relied upon for satisfying regulatory requirements and fundamental safety functions for radionuclide retention. Finally, a set of barrier options are provided so that future economic analyses can select the most viable design path.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2013

Affordable Development and Qualification Strategy for Nuclear Thermal Propulsion

Nuclear Thermal Propulsion (NTP) is a concept which uses a nuclear reactor to heat a propellant to high temperatures without combustion and can achieve significantly greater specific impulse than chemical engines. NTP has been considered many times for human and cargo missions beyond low earth orbit. A lot of development and technical maturation of NTP components took place during the Rover/NERVA program of the 60's and early 70's. Other NTP programs and studies followed attempting to further mature the NTP concept and identify a champion customer willing to devote the funds and support the development schedule to a demonstration mission. Budgetary constraints require the use of an affordable development and qualification strategy that takes into account all the previous work performed on NTP to construct an existing database, and include lessons learned and past guidelines followed. Current guidelines and standards NASA uses for human rating chemical rocket engines is referenced. The long lead items for NTP development involve the fuel elements of the reactor and ground testing the engine system, subsystem, and components. Other considerations which greatly impact the development plans includes the National Space Policy, National Environmental Policy Act, Presidential Directive/National Security Council Memorandum #25 (Scientific or Technological Experiments with Possible Large-Scale Adverse Environmental Effects and Launch of Nuclear Systems into Space), and Safeguards and Security. Ground testing will utilize non-nuclear test capabilities to help down select components and subsystems before testing in a nuclear environment to save time and cost. Existing test facilities with minor modifications will be considered to the maximum extent practical. New facilities will be designed to meet minimum requirements. Engine and test facility requirements are based on the driving mission requirements with added factors of safety for better assurance and reliability. Emphasis will be placed on small engines, since the smaller the NTP engine, the easier it is to transport, assemble/disassemble, and filter the exhaust during tests. A new ground test concept using underground bore holes (modeled after the underground nuclear test program) to filter the NTP engine exhaust is being considered. The NTP engine system design, development, test, and evaluation plan includes many engine components and subsystems, which are very similar to those used in chemical engines, and can be developed in conjunction with them Other less mature NTP engine components and subsystems (e.g., reactor) will be thoroughly analyzed and tested to acceptable levels recommended by the referenced standards and guidelines. The affordable development strategy also considers a prototype flight test, as a final step in the development process. Preliminary development schedule estimates show that an aggressive development schedule (without much margin) will be required to be flight ready for a 2033 human mission to Mars.

Gerrish, Harold P., Jr.↗
NASA NTRS2013

Affordable Development and Qualification Strategy for Nuclear Thermal Propulsion

Nuclear Thermal Propulsion (NTP) is a concept which uses a nuclear reactor to heat a propellant to high temperatures without combustion and can achieve significantly greater specific impulse than chemical engines. NTP has been considered many times for human and cargo missions beyond low earth orbit. A lot of development and technical maturation of NTP components took place during the Rover/NERVA program of the 60's and early 70's. Other NTP programs and studies followed attempting to further mature the NTP concept and identify a champion customer willing to devote the funds and support the development schedule to a demonstration mission. Budgetary constraints require the use of an affordable development and qualification strategy that takes into account all the previous work performed on NTP to construct an existing database, and include lessons learned and past guidelines followed. Current guidelines and standards NASA uses for human rating chemical rocket engines is referenced. The long lead items for NTP development involve the fuel elements of the reactor and ground testing the engine system, subsystem, and components. Other considerations which greatly impact the development plans includes the National Space Policy, National Environmental Policy Act, Presidential Directive/National Security Council Memorandum #25 (Scientific or Technological Experiments with Possible Large-Scale Adverse Environmental Effects and Launch of Nuclear Systems into Space), and Safeguards and Security. Ground testing will utilize non-nuclear test capabilities to help down select components and subsystems before testing in a nuclear environment to save time and cost. Existing test facilities with minor modifications will be considered to the maximum extent practical. New facilities will be designed to meet minimum requirements. Engine and test facility requirements are based on the driving mission requirements with added factors of safety for better assurance and reliability. Emphasis will be placed on small engines, since the smaller the NTP engine, the easier it is to transport, assemble/disassemble, and filter the exhaust during tests. A new ground test concept using underground bore holes (modeled after the underground nuclear test program) to filter the NTP engine exhaust is being considered. The NTP engine system design, development, test, and evaluation plan includes many engine components and subsystems, which are very similar to those used in chemical engines, and can be developed in conjunction with them Other less mature NTP engine components and subsystems (e.g., reactor) will be thoroughly analyzed and tested to acceptable levels recommended by the referenced standards and guidelines. The affordable development strategy also considers a prototype flight test, as a final step in the development process. Preliminary development schedule estimates show that an aggressive development schedule (without much margin) will be required to be flight ready for a 2033 human mission to Mars.

Gerrish, Harold P., Jr.↗
DOE OSTI2024

AGR-5/6/7 CAPSULE 1 THERMAL MODEL WITH OFFSET GAS GAPS

Fuel compact temperatures are a crucial factor in assessing the irradiation performance of tristructural isotropic (TRISO) fuel particles. In the absence of direct measurement, fuel compact temperatures were calculated using a 3 D finite-element thermal model that was subject to simulation uncertainty. The most dominant factor in the uncertainty of calculated fuel temperatures is the gas gap uncertainty caused by the graphite holder nub-to-capsule clearance. This thermal model was revised to examine the most probable graphite offset position for Capsule 1 for four different days during the irradiation. In the analysis, the offset distance and azimuthal direction at both the top and bottom of the holder were varied. The best-fit offset was estimated based on the minimum root mean square error (RMSE) of the residuals (measured minus calculated thermocouple [TC] temperatures) for the operational TCs in Capsule 1. Near the end of the irradiation, when the graphite had shrunk and the gas gaps had grown, a peak temperature was found with maximum offset along the TC lines, with nickel introducing a potential mechanism for fuel particle failures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Investigations about iso-geometric analysis for self-shielding calculations with the subgroup method

The implementation of a self-shielding method for a neutron transport calculation code based on the iso-geometric analysis (IGA) method that can resolve the multi-group neutron transport equation for arbitrary spatial domain, is presented. The self-shielding model based on the subgroup theory is adopted because the subgroup method can be used to perform calculations for any arbitrary geometrical domain which is the main purpose of our IGA code. Some basic theory of the subgroup method is given. A self-shielding calculation based on the PWR fuel pin composed of MOX fuel is presented and compared with a Monte Carlo calculation. The result is that the combination of SN transport theory, IGA and the subgroup method gives correctly shielded cross sections. Validation of the combination of the IGA method and SN neutron transport for subgroup calculations in two steps are presented: the first step is to ascertain the correctness of the IGA solutions compared to a known analytical solution in diffusion theory; the second step is a validation of the IGA solutions compared to a reference calculation in SN transport theory. The relation between the settings for the IGA solver and the accuracy of the results are elucidated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Used Nuclear Fuel Management Using the Next Generation System Analysis Model

The U.S. Department of Energy (DOE) is leading the National effort to manage the back end of the nuclear fuel cycle, encompassing the safe transportation, storage/staging, and/or eventual disposal of used nuclear fuel (UNF) and high-level radioactive waste. The Next Generation System Analysis Model (NGSAM) is DOE’s discrete-event, agent-based simulation tool designed to model the full life cycle of UNF from reactor discharge to final disposal. NGSAM supports the DOE Office of Spent Fuel and High-Level Waste Disposition by enabling a detailed, scenario-based analysis of logistics, infrastructure, and shipping strategies. NGSAM replaces legacy models with a modern, flexible platform built on Repast Simphony and enhanced by the Process Analysis Tool. NGSAM simulates the movement and interaction of individual fuel assemblies with system components such as canisters, casks, railcars, and facilities. The model integrates with the Java Transportation Operations Model to plan and execute transportation scenarios, supporting both constrained and unconstrained resource allocation. Key features include customizable allocation and acceptance algorithms, detailed facility-level operations, and a Quick Edit tool for rapid scenario adjustments. NGSAM supports multimodal transportation modeling (e.g. rail, road, barge) and provides comprehensive cost, schedule, and infrastructure data. NGSAM utilizes data from sources such as DOE’s STANDARDS UNF database and DOE’s Stakeholder Tool for Assessing Radioactive Transportation, while also allowing user-defined inputs for scenario customization. NGSAM enables stakeholders to evaluate complex UNF management strategies, assess system performance under varying assumptions, and inform decision making for future infrastructure investments. Its modular architecture and integration with other Integrated Waste Management System tools make it a critical asset for planning the safe and efficient disposition of the Nation’s growing UNF inventory.

Craig, Brian [Argonne National Laboratory (ANL)]↗
DOE OSTI2021

Evaluation of Engineered Barrier Systems FY21 Report

This report describes research and development (R&D) activities conducted during fiscal year 2021 (FY21) specifically related to the Engineered Barrier System (EBS) R&D Work Package in the Spent Fuel and Waste Science and Technology (SFWST) Campaign supported by the United States (U.S.) Department of Energy (DOE). The R&D activities focus on understanding EBS component evolution and interactions within the EBS, as well as interactions between the host media and the EBS. A primary goal is to advance the development of process models that can be implemented directly within the Generic Disposal System Analysis (GDSA) platform or that can contribute to the safety case in some manner such as building confidence, providing further insight into the processes being modeled, establishing better constraints on barrier performance, etc.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Radiolytic transformation of AHA under single-cycle conditions

Historically, the radiation robustness of extractant ligands and the impact of their radiolysis products have been used to evaluate the radiolytic feasibility of solvent system formulations for used nuclear fuel (UNF) reprocessing. However, other key additives (e.g., reducing and hold-back reagents) are employed by these solvent systems. These chemical species have typically been overlooked from a radiolytic perspective, and consequently little is known on how they impact process performance. Therefore, to holistically support the development of advanced, simplified, single-cycle flowsheets for UNF reprocessing, the radiation robustness of two key proposed additives – (i) the complexing reductant acetohydroxamic acid (AHA) and (ii) the hold-back reagent 1,2-cyclohexylenedinitrilotetraacetic acid (CDTA) – have been evaluated under envisioned process conditions using a combination of steady-state gamma and time-resolved pulsed electron irradiation techniques.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Exascale Multiphysics Nuclear Reactor Simulations for Advanced Designs

ENRICO is a coupled application developed under the U.S. Department of Energy's Exascale Computing Project (ECP) targeting the modeling of advanced nuclear reactors. It couples radiation transport with heat and fluid simulation, including the high-fidelity, highresolution Monte-Carlo code Shift and the Computational fluid dynamics code NekRS. NekRS is a highly-performant open-source code for simulation of incompressible and low-Mach fluid flow, heat transfer, and combustion with a particular focus on turbulent flows in complex domains. It is based on rapidly convergent high-order spectral element discretizations that feature minimal numerical dissipation and dispersion. State-of-the-art multilevel preconditioners, efficient high-order time-splitting methods, and runtime-adaptive communication strategies are built on a fast OCCA-based kernel library, libParanumal, to provide scalability and portability across the spectrum of current and future high-performance computing platforms. On Frontier, Nek5000/RS has recently achieved an unprecedented milestone in breaching over 1 billion spectral elements and 350 billion degrees of freedom. Shift has demonstrated the capability to transport upwards of 1 billion particles per second in full core nuclear reactor simulations featuring complete temperature-dependent, continuous-energy physics on Frontier. Shift achieved a weak-scaling efficiency of 97.8% on 8192 nodes of Frontier and calculated 6 reactions in 214,896 fuel pin regions below 1% statistical error yielding first-of-a-kind resolution for a Monte Carlo transport application.

Hamilton, Steven P.↗
DOE OSTI2023

Sensitivity Analysis of Transient Critical Heat Flux by RIA Under High-Pressure Flow Boiling Conditions in TRTL

A reactivity-initiated accident (RIA) is a design-basis accident under which critical heat flux (CHF) is likely to be exceeded. The operational margin for RIAs is currently determined using steady-state CHF lookup tables, which provide conservative estimates relative to transient CHF phenomena. The Transient Reactor Test Loop (TRTL) facility at Oregon State University is capable of performing out-of-pile rapid heating experiments representative of a RIA at conditions representative of a pressurized water reactor (PWR). Here, to further our understanding of and ability to predict transient CHF under PWR conditions, we performed a sensitivity analysis on a RELAP5-3D model of the TRTL facility coupled to the RAVEN code framework to define a proposed experimental test matrix to be performed at the TRTL facility. We then implemented a flow boiling CHF correlation into RELAP5-3D and performed a secondary sensitivity analysis inspecting the impact of the built-in RELAP5-3D CHF and heat transfer multipliers on both the prediction of CHF and key safety parameters, such as peak cladding temperature and heat flux. The results show that the multiplier with the highest influence toward the prediction of CHF occurrence and the safety parameters is the transient CHF multiplier. Operational performance envelopes have been developed for each of the test matrix cases and will be used for validation once the experiments are performed. The TRTL facility is currently performing shakedown testing to verify system performance prior to proceeding with the experimental campaign. Restart testing results include pump curve restart testing, pressure tests, and heater rod thermocouple transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Review of thermophysical property methods applied to fueled and un-fueled molten salts

Recent investigation into designs of molten salt reactors (MSR) requires accurate data of candidate salts. Review of the literature shows lacking thermophysical (TP) property data for coolants and fuel-bearing salts of Advanced High Temperature Reactors (AHTR). Measurement uncertainties of current data are even more deficient, and this affects ongoing research where MSR modeling and testing are being performed. Melting temperature, viscosity, density, specific heat capacity, and thermal and electrical conductivity measurements are often not arranged together or discussed in great detail, especially for fuel-bearing salts. Raw data from key findings spanning over 70 years for LiF, LiF-BeF 2 , LiF-NaF-KF, KCl-MgCl 2 , KF-ZrF 4 , and NaF-KF-MgF 2 have been organized and presented here. Brief description of methods used to obtain these results, in addition to the TP data of some of these salts as fuel carriers, has also been discussed. Thermal conductivity data are particularly weak due to older, high-error measurement techniques. More recent measurement systems to evaluate salt thermal conductivity show promise for obtaining consistent absolute values. This work concludes that measurement systems such as transient hot-wire method, forced Rayleigh scattering, and laser flash analysis provide the most promise to fill these data gaps and reduce current uncertainty in thermal conductivity measurements of both fueled and un-fueled molten salts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Multiphysics Modeling of Microreactors with NEAMS codes, and Validation Based on KRUSTY Reactivity Insertion

The NEAMS Multiphysics Applications team continues to assess code usability and functionality for microreactor design and safety analyses, while demonstrating that NEAMS tools capture both steady-state and transient behavior across distinct microreactor concepts. In FY2025, the team advanced full-core, high-fidelity, multiphysics models that solve more complex problems and strengthen verification/validation for several microreactor systems: heat-pipe microreactor (HPMR), gas-cooled microreactor (GCMR), and the KRUSTY experiment. These models employ the MOOSE MultiApp/Transfers architecture with Griffin for neutronics, BISON for heat conduction/thermomechanics, Sockeye for heat pipes, SAM/THM for coolant channels and loops, and SWIFT for hydride behavior, with meshes generated via the MOOSE Reactor Module. The graphite models available in the Grizzly code were also investigated for future analyses. For the HPMR, a Na-HPMR variant was constructed to align with recently validated heat-pipe experiments and Sockeye’s LCVF capability, enabling mechanistic heat-pipe transients and startup modeling. The Na-HPMR will serve as the primary model for HPMR investigations in upcoming tasks. The load-following and single heat-pipe failure scenarios (Griffin/BISON/Sockeye), which were previously modeled for the K-HPMR, were replicated for the Na-HPMR, showing strong negative temperature feedback and highly localized thermal effects, respectively, while the startup case captured vapor-front progression and heat-removal activation. Solid mechanics was added to the previously built K-HPMR full-core model in BISON, showing minimal impact on steady-state reactivity yet enabling stress-field predictions that prepare the path for full-core TRISO performance analyses. For the GCMR, automated steady-state and four transient scenarios were executed using Griffin/BISON/SAM/SWIFT. Results confirm robust inherent safety: power collapses promptly in loss-of-cooling events, the inlet-temperature drop settles to a new equilibrium, and a single-channel blockage yields only a ~30 K local fuel-temperature rise with <0.4% power decrease. SWIFT-predicted hydrogen redistribution affects reactivity during both steady-state and transient conditions, underscoring its importance. A Brayton-cycle balance of plant (BOP) model in SAM/THM demonstrated stable startup behavior, and xenon-driven reactivity during load following was analyzed. To improve TRISO-compact temperature fidelity, a fast multiscale Heat Source Decomposition (HSD) treatment was implemented. Against heterogeneous benchmarks, HSD reduces underprediction of kernel temperatures and lowers predicted peak powers in reactivity-insertion transients compared to previous homogenized models. KRUSTY warm-critical validation progressed from FY2024 baselines: the 15Ȼ insertion shows excellent agreement in peak power (~2% high) and temperature trends, and the 30Ȼ case was automated via a feedback controller that maintained power near 3 kW for ~150 s with close agreement to data. The successful modeling of the warm critical tests has laid a strong foundation for simulating more complex nuclear system tests in the years ahead. Throughout FY2025, developer feedback was provided (e.g., MOOSE batch mesh generation, distributed pre-split meshes, Griffin sweeper on displaced meshes), several new models were contributed to the Virtual Test Bed, and an OECD-NEA WPRS multiphysics benchmark based on the HPMR was initiated to enable broader cross-comparison and best-practice development with the nuclear community at large.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗