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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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At least 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

3D TRISO particle-explicit compact meshing

The TRI-structural ISOtropic (TRISO) layered fuel particle is a robust nuclear fuel form offering enhanced safety and performance for advanced reactor concepts, including high-temperature gas-cooled reactors and other Generation IV designs. These poppy-seed-sized particles are embedded in a graphite matrix to form fuel elements that must withstand elevated temperatures and high burn-up levels. The heterogeneous nature of these fuel elements — comprising thousands of randomly distributed TRISO particles — produces complex stress fields and thermal gradients that one- and two-dimensional models cannot accurately capture. While three-dimensional modeling has improved predictions of dimensional changes, internal pressure buildup, and fission product transport under irradiation, current approaches rely on homogenized material properties that are known to have considerable divergence from experimental observations. This work presents a methodology for optimized random packing of TRISO fuel compacts and full three-dimensional mesh generation within the BISON fuel performance code, with each particle coating layer individually discretized. The resulting mesh was demonstrated through heat conduction simulations under representative in-reactor operating conditions, showing strong agreement with expected behavior. This capability enables detailed analysis of particle-to-particle interactions, matrix cracking mechanisms, and the statistical distribution of coating layer failures — all of which directly govern fuel performance and safety margins.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Process Heat for Chemical Industry

Process heating with nuclear energy can reduce greenhouse gas emissions by reducing combustion of fossil fuels in fired heaters as steam boilers. Light water reactors can replace the majority of steam duties used by industry; however, high temperature processes such as steam methane reforming require advanced high temperature reactors. Here, guidance on matching the scale of nuclear reactors with specific industries is provided. Principles of heat transport, temperature boosting, and substitution for hot combustion gases are also discussed in this section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Comparative study with two reduced test facilities for air-cooled RCCS scaling law

Korea Atomic Energy Research Institute (KAERI) and Argonne National Laboratory (ANL) conducted steady-state tests at constant heat flux conditions to investigate the scale effect on heat removal behavior of the air-cooled Reactor Cavity Cooling System (RCCS). Two differently scaled-down test facilities were used in this study, KAERI’s ¼ scale with 4.0-m cavity height, and ANL’s ½ scale with 6.8-m cavity height. Two scaling laws were proposed to simulate the radiation across the cavity and the buoyancy-driven duct flow in the riser, respectively. The test matrix for KAERI and ANL test facilities focused on the scaling effect of the heated riser length with constant heat flux at the PMR200 RCCS design. The test results showed that mixed convection in the riser duct is an important factor for accurately extrapolating the thermo-fluid behavior in the prototype from the test results in the scale-down facilities. The system analysis code, GAMMA+, with improved heat transfer models predicted fairly well the air-cooled RCCS test data from two facilities. GAMMA+ analysis results showed that the predicted radiation fractions on the heated plate in the scale-down test conditions were larger than those on the reactor vessel in the prototype. The scaling law for air-cooled RCCS was improved by considering the mixed convection in the riser duct and the same radiation fraction on the heated plate. The mixed convection effect was calculated by the height ratio. The same radiation fraction provided a conservative extrapolation from the scale-down test results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Effects of break geometry and orientation on helium-air mixing in simulated reactor cavities of high temperature gas reactors

Here, this study experimentally examined the spatial and temporal variations in air and helium concentrations and temperature fields within simulated reactor cavities of a High Temperature Gas Reactor (HTGR) following helium discharge into an initially air-filled reactor cavity system. Detailed temperature maps were generated using a combination of fiber optics temperature sensor and multiple thermocouple probes within the simulated reactor cavities. The research scenario involved a hypothetical small pipe break in the Reactor Pressure Vessel, resulting in the release of high-temperature helium into the surrounding cavity. A scaled multi-compartment experimental facility, modeled after the General Atomics Modular High Temperature Gas Reactor (GA-MHTGR) design, was constructed for helium and air mixing experiments. Oxygen sensors and thermocouple probes were installed in all five cavities to measure the concentrations of oxygen (or helium) and the temperature distributions of the gas mixture. The experimental findings highlighted the significant impact of the injected helium jet velocity on the gas mixing process and demonstrated how the direction of the helium jet influences the air-helium temperature profiles within the cavities.

Air-ingress↗
DOE OSTI2023

TRISO Burnup-Dependent Failure Analysis of a HTGR Design-Basis Accident Using BISON

Here, this work assesses the failure behavior of the silicon carbide (SiC) layer in TRIstructural ISOtropic (TRISO) fuel particles with BISON during both steady-state and transient conditions for the MHTGR-350 design by General Atomics. A one-dimensional BISON model for a uranium oxycarbide–bearing TRISO is developed to simulate a power level of 50 mW/particles up to 12.4% fissions per initial metal atom (FIMA). Stress predictions for the materials of interest are presented as a function of burnup, along with the SiC failure probability computed using Weibull statistics under the assumption of realistic SiC quality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Performance Comparison of Helium and sCO 2 as Coolants for Modular Finger-Type Divertors

Several studies at the Georgia Institute of Technology have evaluated the thermal and fluids performance of the helium-cooled modular divertor with multiple jets (HEMJ) over the past decade. This finger-type divertor was studied both experimentally at nearly prototypical conditions and numerically at fully prototypical operating conditions using experimentally validated simulations. Recently, supercritical carbon dioxide (sCO 2 ) has been studied as the primary coolant in power cycles and other applications in various systems, in part because CO 2 achieves the high densities typical of supercritical fluids at relatively low temperatures and pressures, with a critical point of (7.38 MPa, 31°C). This density makes it possible to realize very compact and efficient sCO 2 power cycles. The feasibility of sCO 2 as a coolant for plasma-facing components, specifically the divertor, was therefore evaluated as part of the Fusion Energy System Studies design study activities. This work compares the thermal-fluid performance of helium and sCO 2 in the HEMJ divertor geometry using numerical simulations at prototypical conditions: inlet temperatures T i = 600°C to 700°C, pressures p ≈ 10 MPa, and steady-state incident heat fluxes on the tile q″ < 17 MW/m 2 . The performance is quantified here as the maximum heat flux that can be accommodated by the plasma-facing tile, the pumping power fraction, defined as the ratio of the coolant pumping power to the incident thermal power, and the operating stress limits based on ASME pressure vessel criteria. As expected, helium requires lower mass flow rates and pumping power fractions within imposed maximum temperature limits for the HEMJ pressure boundary. However, it also appears that neither helium nor sCO 2 can remove 10 MW/m 2 of incident heat flux while meeting ASME pressure vessel criteria. Finally, the numerical modeling reveals that sCO 2 may remove slightly higher incident heat fluxes than helium due to the imposed stress limits due to the sCO 2 coolant resulting in smaller local temperature gradients, albeit at a considerably higher pumping power fraction.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Horizontal Split Table Conceptual Design for Validation of Nuclear Data used in Advanced Reactors [Slides]

A methodology to create conceptual designs of benchmark critical experiments for advanced reactors nuclear data testing and validation was developed. A first concept was explored, pebble-bed high-temperature gas cooled reactor, based on the HTR-10 reactor. The very high correlation is a proof of concept that the design is similar to the application, and performing such critical experiments would help nuclear data testing and validation. This proof-of-concept was included in the IER-539 CED-1:Preliminary Design of a New Horizontal Split Table report. Other concepts could be explored if needed, such as Molten-salt reactor, Sodium-cooled fast reactor or Heat pipe reactors/Microreactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Finite Volume Discretization of the Euler Equations in Pronghorn

Modeling flow and heat transfer in high temperature gas reactors (HTGR) requires the ability to model a wide range of flow speeds from slow (natural convection), to intermediate (forced-flow conditions), to supersonic regimes (depressurization) for a wide range of geometries including the pebble bed, upper and lower plenum, and risers. In previous work, Pronghorn has effectively modeled low-to-medium speed flows in scenarios such as the one described in the two-dimensional PBMR-400 benchmark, using its finite-element-based streamline-upwind Petrov-Galerkin (SUPG) stabilized implementation of the Euler equations. However, limitations of this method become apparent when dealing with more complicated geometries (e.g. imposing slip boundary conditions at nodes belonging to two different boundaries) and when gas speeds are fast enough for shocks and supersonic flow to occur. For these problems, the finite-element-based solver lacks robustness and is plagued by slow iterative convergence or even divergence. In order to address these challenges, the Pronghorn code at INL has been updated with new, modified versions of its original equations. The new Pronghorn models are built on the finite volume method with a Harten-Lax-van Leer-Contact (HLLC) Riemann solver based numerical flux method, which (1) allows imposing slip boundary conditions much more robustly and (2) performs well for a wide range of flow speeds. The finite-volume-based flow solver will form the basis for a robust coarse-mesh thermal-hydraulics capability in Pronghorn.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Investigation of Point-Contact Strategies for CFD Simulations of Pebble-Bed Reactor Cores

This study numerically investigated the effects of various contact strategies on the thermal hydraulic behavior within a structured bed of 100 explicitly modeled pebbles. Four contact strategies and two thermal hydraulic conditions were considered. The strategies to avoid contact singularities include decreasing the pebble diameter, increasing the pebble diameter, bridging the pebble surfaces near the contact region, and capping the pebble surfaces near the contact region. One strategy, Strategy 3a, which involves bridging with a cylinder equal to 10% of the pebble diameter, was selected as the baseline strategy because it addressed the contact singularity while minimizing the geometric changes that affect the bed porosity. The two thermal hydraulic conditions were full-power operation (Case 1) and pressurized loss of forced cooling or PLOFC (Case 2). Simulations of the conjugate heat transfer within the structured bed were performed using the Reynolds-averaged Navier–Stokes approach with the realizable k-ϵ turbulence model and two-layer all y + wall treatment. The thermal-fluid quantities of interest were compared between the contact strategies for each case. In Case 1, the hydraulic behavior was sensitive to the contact strategy, with large differences in the pressure drop (30%) and volume-average velocity (4%). The thermal behavior was not sensitive, with less than a 0.5% difference across the strategies. To better understand the separate effects of each heat transfer mode, Case 2 was divided into the following subcases: conduction (Case 2a); conduction/radiation (Case 2b); and conduction/radiation/convection (Case 2c). Case 2a represents an early phase of the PLOFC transient. Case 2b represents an intermediate phase of the PLOFC transient, with the pebble temperatures sufficiently high for the radiative heat transfer to be non-negligible. Case 2c represents a late phase of the PLOFC transient after the establishment of the natural circulation of the heat transfer fluid. For Case 2, large differences in the contact strategy were observed only in Case 2a with only conduction. The difference in the maximum pebble temperature was 23% in Case 2a, 2% in Case 2b, and 0.3% in Case 2c.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRSDate not supplied

Coupled Reactor Multiphysics and Mass Scalability Assessment for Crewed Megawatt-Class NEP System Architectures

Nuclear Electric Propulsion (NEP) is an in-space propulsion technology capable of enabling opposition and conjunction class crewed Mars missions. NEP subsystems include the reactor for heat generation, a power conversion system (PCS), power management and distribution (PMAD), electric propulsion subsystem (EPS), and a primary heat rejection system. Specific mass, or αe (kg / kWe), is a key performance parameter (KPP) of the propulsion system which is directly scalable with the performance and mass estimates for individual components. To inform technology maturation planning, full system and component level parametric modeling is ongoing to explore the design trade space and illustrate the effect of subsystem design choices on the system KPPs. In this study, scaling of high-assay, low-enriched uranium (HALEU) reactor designs is assessed through coupled reactor physics and thermal hydraulics analyses. Scaling analyses evaluate the impact of system performance parameters (power level, interface temperatures) on mass for direct gas cooled, pumped liquid metal, and passively cooled heat pipe reactor concepts. Each concept requires specific geometries and working fluids to reach the performance goals of PCS interface conditions (temperature, pressure, flow rate) and system mass. The reactor assembly includes the active core (fuel, moderator, cladding, working fluid), axial and radial neutron reflectors, control drums, structural support / pressure vessel, and external radiation shielding. Each of these components are parametrically sized based on performance parameters for a megawatt-class power cycle. Results of this scaling analysis increase NEP propulsion system modeling fidelity and ultimately aim to support technology down-selection along with related technology development planning. The reactor and shield αe are a function of several PCS design choices, and reactor scaling with these parameters must be considered to enable an informed decision on reactor geometry and working fluid combination.

Nuclear Electric Propulsion↗
NASA NTRSDate not supplied

Coupled Reactor Multiphysics and Mass Scalability Assessment for Crewed Megawatt-Class NEP System Architectures

Nuclear Electric Propulsion (NEP) is an in-space propulsion technology capable of enabling opposition and conjunction class crewed Mars missions. NEP subsystems include the reactor for heat generation, a power conversion system (PCS), power management and distribution, electric propulsion system, and heat rejection system. Specific mass, or α (kg/kWe), is a key performance parameter (KPP) of the propulsion system which is directly scalable with the performance and mass predictions for each individual component. To inform technology maturation planning activities, full system and component level parametric modeling is ongoing to explore the design trade space and illustrate the effect of subsystem design choices on the system KPPs. In this study, scaling of high-assay, low-enriched uranium reactor designs is assessed through coupled reactor physics and thermal hydraulics analyses. Scaling analyses evaluate the impact of system performance parameters (power level, interface temperatures) on mass for direct gas cooled, pumped liquid metal, and passively-cooled heat pipe reactor concepts. Each concept requires specific geometries, fluids, and power conversion interface conditions (temperature, pressure, flow rate) to meet desired performance and mass. The reactor assembly includes the active core (fuel, moderator, cladding, working fluid), axial and radial neutron reflectors, control drums, structural support / pressure vessel, and external radiation shielding. Each of these components are parametrically sized based on performance parameters for a megawatt-class power cycle. Results of this scaling analysis increase NEP propulsion system modeling fidelity and ultimately aim to support concept down-selection along with related technology development planning. The reactor and shield α are a function of several PCS and heat rejection system design choices, and reactor scaling with these parameters must be considered to enable an informed decision on an optimal reactor geometry and working fluid combination.

Nuclear Electric Propulsion↗
DOE OSTI2020

DOE-ART HT(G)R R&D Overview

The US Government currently has no plans to finance or build advanced nuclear reactors. Instead, the DOE performs research and development to raise the technological readiness of new concepts, fuels and component materials so that private industrial partners are more likely to design and build plants. Develop technologies that can enable new concepts and designs to achieve greater levels of safety and resilience, flexibility of use, sustainability and construction or operational affordability. Collaborate with industry to identify and conduct essential research to reduce technical risk associated with advanced reactor technologies. Sustain essential technical expertise and capabilities within national laboratories, universities, and industry to perform needed research. Collaborate with the Nuclear Regulatory Commission and Standards Developing Organizations (ASME, ANS, ANSI, etc.) to address gaps in codes and standards to support advanced reactor designs

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Lessons Learned from RELAP5-3D Modeling of HTTF

Modeling of HTTF using the RELAP5-3D ring model has shown an inability to predict transient temperature rise during HTTF experiments. This presentation provides an overview of the RELAP5-3D ring model, the HTTF modeling activities using the ring model, and reasons why the models failed to reproduce measured temperature rise. It also presents an overview of the path forward for RELAP5-3D modeling of HTTF to better predict the behavior of HTTF experiments

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

GIF: VHTR Benchmark

Intern research presentation for C-120 Group meeting.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗