Insights from a Joint US/India Workshop on Modeling and Simulation for Improved Nuclear Material Accountancy.
Abstract not provided.
Shoman, Nathan↗
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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
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The Use of Immersion Rigs for High Temperature Hydrogen Exposure Testing within the Nuclear Thermal Rocket Element Environmental Simulator (NTREES): Thermal Soak Rig (TSR)
The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility was purpose constructed to perform non-nuclear evaluations of nuclear thermal propulsion (NTP) system fuel materials and structures within prototypic thermochemical environments. This system has been utilized steadily in its ability to subject test specimens to thermochemical and thermohydraulic environments simulating that of an operating nuclear rocket engine. Fission heat is simulated by induction power and experiments are conducted within a ~1000 psi pressure vessel. Hydrogen is conventionally passed through the heated fuel surrogate test specimen while pressure, temperature, and gas species data are collected at various points along the experiment. In order to test fuel and material coupon samples, a class of test apparatus named “immersion rigs” are being developed and employed to more rapidly test these smaller and more technically challenging test specimen. One example of a promising potential fuel structure, Tristructural-isotropic (TRISO) particles, presents unique challenges for testing of this type. TRISO fuel micro-particles are spheroids typically on the order of 500 – 1000 μm in diameter, and exposing a batch sample to hot hydrogen requires purpose-built special test equipment. Thusly, an immersion rig was developed and successfully demonstrated to expose ~1 g of micro-particles to hydrogen gas at temperatures and pressures relevant to NTP systems for the purpose of fuel evaluation. The rig, comprised primarily of graphite and pure tungsten, houses in its core a batch of micro-particles between pucks of porous silicon carbide (SiC). This approach permits gas flow while simultaneously retaining the particles in place. Herein is a discussion of the design, analysis, fabrication, and testing of the NTREES Thermal Soak Rig (TSR).
Space Nuclear Propulsion↗
Photopolymerizable Sol-gel Synthesis for Faster Manufacturing of Surrogate Nuclear Debris
Simulated debris is used to train first responders for potential nuclear detonations. Sol gel chemistry has been used to produce surrogate material for these training exercises, but the current average production time is 4 hours. Optimization using UV-polymerizable sol gel precursors can potentially decrease time as well as cost of the production process. FT-IR spectroscopy was used to measure and compare gelation kinetics of the standard and photopolymerizable sol gel synthetic procedures. Addition of the UV-polymerizable precursors and controlling the aqueous phase volume resulted in a 120-fold decrease in vitrification time.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Hypothetical Nuclear Reactor Facility Modeling Simulation Data Scenario Comparison
This document evaluates a hypothetical nuclear powerplant and associated protective force personnel using modern modeling and simulation tools. The facility incorporates security early in the design to consider and integrate methods to resolve security issues and vulnerabilities via the facility’s inherent design characteristics before construction. The evaluation in this document is an example only. It is not intended to recommend or evaluate the effectiveness of existing physical security requirements or identify any method that the U.S. Nuclear Regulatory Commission staff may find acceptable for complying with existing requirements.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Identification of preferred multimodal ligand‐binding regions on IgG1 F C using nuclear magnetic resonance and molecular dynamics simulations
Abstract In this study, the binding of multimodal chromatographic ligands to the IgG1 F C domain were studied using nuclear magnetic resonance and molecular dynamics simulations. Nuclear magnetic resonance experiments carried out with chromatographic ligands and a perdeuterated 15 N‐labeled F C domain indicated that while single‐mode ion exchange ligands interacted very weakly throughout the F C surface, multimodal ligands containing negatively charged and aromatic moieties interacted with specific clusters of residues with relatively high affinity, forming distinct binding regions on the F C . The multimodal ligand‐binding sites on the F C were concentrated in the hinge region and near the interface of the C H 2 and C H 3 domains. Furthermore, the multimodal binding sites were primarily composed of positively charged, polar, and aliphatic residues in these regions, with histidine residues exhibiting some of the strongest binding affinities with the multimodal ligand. Interestingly, comparison of protein surface property data with ligand interaction sites indicated that the patch analysis on F C corroborated molecular‐level binding information obtained from the nuclear magnetic resonance experiments. Finally, molecular dynamics simulation results were shown to be qualitatively consistent with the nuclear magnetic resonance results and to provide further insights into the binding mechanisms. An important contribution to multimodal ligand‐F C binding in these preferred regions was shown to be electrostatic interactions and π–π stacking of surface‐exposed histidines with the ligands. This combined biophysical and simulation approach has provided a deeper molecular‐level understanding of multimodal ligand–F C interactions and sets the stage for future analyses of even more complex biotherapeutics.
Gudhka, Ronak B.↗
Coupled physics simulation of fracture in nuclear fuel pellets induced by resistive heating
While it is well known that the UO 2 fuel used in light-water reactors cracks early in its life in the reactor, it is very difficult to collect in-reactor data regarding fuel cracking initiation and propagation to validate fracture models in fuel performance codes. Most fracture data comes from the post-irradiation examination of fuel that has been subjected to long and complex power histories. Replicating the thermal gradients experienced in this fuel to study crack initiation and propagation in a laboratory environment is challenging. This is primarily because the fuel is heated volumetrically by fission, which is difficult to replicate. One way to approximate this volumetric heating out of the reactor is to use electrical resistance (or Joule heating). An electrical conductivity model has been developed and coupled with the existing thermomechanical modeling capability in the BISON nuclear fuel performance code. This will permit simulation of experiments that use resistive heating for the purpose of validating fracture models in BISON. The modeling approach developed here is applied to simulate both historical and current resistive heating experiments on ceramic nuclear fuel.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Simulation of the nuclear-electromagnetic cascade development in an ionization spectrometer
Longitudinal nuclear-electromagnetic cascade in ionization spectrometer simulated by Monte Carlo method, noting energy relationship to particle sum recording
Jones, W. V.↗
Enabling the Broader Use of MOOSE for Nuclear Energy and Other Simulation
This Final Scientific and Technical Report summarizes work performed under the Phase IIA SBIR project “Enabling the Broader Use of MOOSE for Nuclear Energy and Other Simulation” (DE-SC0020906) from August 2023 through August 2025. The objective of the Phase IIA effort was to mature and harden capabilities developed during Phase II, with the goal of enabling practical interoperability between Coreform’s isogeometric analysis (IGA) technologies and the Multiphysics Object-Oriented Simulation Environment (MOOSE), while improving robustness, performance, and scalability for complex, nuclear-relevant geometries. Over the course of Phase IIA, the project established and validated an extraction-based interoperability pathway between Coreform tools and MOOSE. A combined mesh and matrix format was defined collaboratively with MOOSE developers and integrated into the solver, enabling standard MOOSE workflows to operate on data exported from Coreform’s IGA and Flex Representation Method (FRM) pipelines. Early demonstrations validated architectural compatibility using linear solid mechanics problems, while later efforts focused on benchmark testing and external use. By the end of the project period, engineers at BWXT were able to independently set up and execute a simulation using the Coreform–MOOSE workflow and provide direct feedback that informed further refinement. In parallel, substantial effort was devoted to improving the robustness of trimmed U-spline construction for complex CAD geometries. A growing test suite of nuclear-relevant models was compiled through collaboration with multiple stakeholders and used to drive extensive bug fixing and reliability improvements. These efforts resulted in improved robustness and performance, including the addition of fallback capabilities that enhance reliability when the underlying commercial CAD kernel fails. Performance-oriented work progressed later in the project, with the development and demonstration of methods to decompose complex geometries into structured subregions and updated data representations to support more efficient solver processing. Additionally, extensive enhancements to threadsafe parallel data structures and trimming operations established a foundation for scalable processing of large assemblies. Collaboration with Sandia National Laboratories on the SGM geometric modeling kernel advanced to a functioning interface test case, positioning the workflow for future kernel integration. Overall, the Phase IIA effort successfully transitioned the project from architectural proof-of-concept to externally exercised, solver-integrated capability, while clarifying remaining technical challenges related to standardization, performance optimization, and kernel integration.
42 ENGINEERING↗
Thermal Simulator Development: Non-Nuclear Testing of Space Fission Systems
Non-nuclear testing can be a valuable tool in the development of a space nuclear power system. At the NASA MSFC Early Flight Fission Test Facility (EFF-TF), highly designed electric heaters are used to simulate the heat from nuclear fuel to test space fission power and propulsion systems. To allow early utilization, nuclear system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. In this test strategy, highly designed electric heaters are used to simulate the heat from nuclear fuel, allowing one to develop a significant understanding of individual components and integrated system operation without the cost, time and safety concerns associated with nuclear testing.
Bragg-Sitton, Shannon M.↗
Initial Operation and Shakedown of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)
To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Prototypical fuel elements mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission in addition to being exposed to flowing hydrogen. Recent upgrades to NTREES now allow power levels 24 times greater than those achievable in the previous facility configuration. This higher power operation will allow near prototypical power densities and flows to finally be achieved in most prototypical fuel elements.
Emrich, William J., Jr.↗
The dissolution of simulant vitrified intermediate level nuclear waste in young cement water
Abstract It is pertinent to the safety case for geological disposal in the UK that the behaviour of vitrified wastes in proximity to cementitious materials is understood. In this study, vitrified simulant intermediate level nuclear waste (ILW) was subject to dissolution in a synthetic cement water solution to simulate disposal conditions. Results show that the presence of alkali / alkaline earth elements in the cementitious solution can be favourable, at least in the short-term, leading to lower dissolution rates associated with incorporation of these elements into the altered layer of the glass.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Multiphysics FISPACT-II and TENDL-2019 simulation: neutron-induced damage metrics
Nuclear interactions can be the source of atomic displacement, irradiation-induced defects and transmutation in structural materials. Such quantities are derived from, or can be correlated to, nuclear kinematic simulations of primary atomic energy distributions spectra and the quantification of the numbers of secondary defects produced per primary as a function of the available recoils, residual and emitted, energies. Recoil kinematics of neutral, residual, gas, proton, alpha particle emissions are now more rigorously treated based on recent, complete and enhanced nuclear data parsed in state-of-the-art processing tools. Defect production metrics are the starting point in the complex problem of correlating and simulating the behaviour of materials under irradiation, as experimental information is rare or scattered. Detailed, segregated primary knock-on-atom metrics are now becoming available as the starting point of further simulation processes of isolated and clustered defects in material lattices. This allows more materials, lattices, neutron incident energy ranges, and irradiation conditions to be explored with sufficient data to adequately cover both standard and novel applications and materials: the broader reactor applications landscape. The damage metrics of of materials are systematically explored under typical but different reactor's type environment. The inventory code FISPACT-II combined with the enhanced nuclear data forms of the TENDL-2019 libraries allow one to not only calculate dpa from mostly scattering events but to also properly predict gas production, nuclear heating and transmutation under the same conditions.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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↗
Scalable algorithms for domain decomposition of Monte Carlo neutral particle transport simulations on unstructured mesh
Monte Carlo simulations are the gold standard in particle transport simulations, for shielding and nuclear reactor simulations Full core simulations can require terabytes of memory for tracking pin-resolved (up to 50k pins), nuclide (200 tracked), axial (20-200 zones) reaction rates Domain decomposition, splitting the calculation and the memory storage, among several nodes is challenging as it involves communication of particles across domain boundaries We created a new MOOSE application, MaCaw, to study this problem Algorithms developed for ray tracing in MOOSE were adapted for simulating neutral particle transport Physics and tally routines in OpenMC, dynamically linked, are called from the application
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Nuclear Thermal Rocket Element Environmental Simulator (NTREES)
A detailed description of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) is presented. The contents include: 1) Design Requirements; 2) NTREES Layout; 3) Data Acquisition and Control System Schematics; 4) NTREES System Schematic; and 5) NTREES Setup.
Schoenfeld, Michael↗
Climatic Effects of Regional Nuclear War
We use a modern climate model and new estimates of smoke generated by fires in contemporary cities to calculate the response of the climate system to a regional nuclear war between emerging third world nuclear powers using 100 Hiroshima-size bombs (less than 0.03% of the explosive yield of the current global nuclear arsenal) on cities in the subtropics. We find significant cooling and reductions of precipitation lasting years, which would impact the global food supply. The climate changes are large and longlasting because the fuel loadings in modern cities are quite high and the subtropical solar insolation heats the resulting smoke cloud and lofts it into the high stratosphere, where removal mechanisms are slow. While the climate changes are less dramatic than found in previous "nuclear winter" simulations of a massive nuclear exchange between the superpowers, because less smoke is emitted, the changes seem to be more persistent because of improvements in representing aerosol processes and microphysical/dynamical interactions, including radiative heating effects, in newer global climate system models. The assumptions and calculations that go into these conclusions will be described.
Oman, Luke D.↗
Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)
To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). Last year NTREES was successfully used to satisfy a testing milestone for the Nuclear Cryogenic Propulsion Stage (NCPS) project and met or exceeded all required objectives.
Emrich, William J., Jr.↗
ONIX: An open-source depletion code
Open Source software enables innovative, community-based software development. ONIX brings this concept to the field of depletion calculations. It is an open-source depletion software to be used for nuclear reactor simulations, for fissile material production analysis as well as for nuclear arms control applications. ONIX provides a module to solve the depletion equation using a Chebyshev Rational Approximation Method. For the generation of one-group cross sections, it includes a coupling interface for the open-source neutron transport code, OpenMC, as well as a module to read pre-computed values in a stand-alone mode. ONIX has special features to optimize nuclear data libraries, to update isomeric branching ratio during burnup, and to support automation of simulations for nuclear archaeology. In conclusion, ONIX has been validated against results from numerical and experimental benchmarks, and its results agree with other methods within expected error ranges.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗