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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 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

The Monolithic Heat Pipe Microreactor Reference Plant Model

This work introduces a reference plant model for a generic monolithic heat-pipe-cooled microreactor. The model will serve as a springboard to develop future evaluation models in the licensing process of similar microreactor designs at the U.S. Nuclear Regulatory Commission. This model has been developed with the Comprehensive Reactor Analysis Bundle and its specifications are based on open literature publications for the eVinci TM design. BlueCRAB is the U.S. Nu- clear Regulatory Commission non-light-water reactor analysis system based on the Multiphysics Object-Oriented Simulation Environment framework, which can couple the Griffin, BISON, and Sockeye applications to resolve the various physics that are essential for the safety analysis of this type of reactor system. The core specifications includes tristructural isotropic fuel, graphite monolith, graphite reflectors, and drums composed of graphite and B 4 C. No moderator or burnable poison pins are used in the design. The fuel enrichment is reduced to control excess reactivity in the core. This core design is not optimized and only serves for testing purposes, since the primary objective of this work is to exercise the multiphysics coupling for this type of reactor system. A three dimensional (3D) core heterogeneous Griffin discrete ordinates (SN) transport model allows the precise calculation of the flux distribution and pin powers. Griffin transfers the power density distribution and obtains a temperature distribution to and from BISON. The BISON model com- putes the 3D core temperature distribution and is coupled to 876 Sockeye subapplications running a heat pipe model. This 3D conduction model is coupled to the various heat pipes via heat flux boundary conditions. The model includes a small gap between the heat pipe and the monolith. Convective heat transfer boundaries with either ambient temperature or condenser temperature as heat sinks are imposed at the model boundaries. The 2D Sockeye heat pipe model uses a vapor- only methodology, which provides the needed resolution for transient calculations and allows the determination of various heat pipe limits. This approach is superior to the superconductor model traditionally used in steady-state calculations. BlueCRAB computes steady-state power and temperature distributions that serve as the initial condition for a loss-of-heat-sink transient simulation. The steady-state results show significant peaking due to the position of the control drum, but this is a characteristic of the particular design used, which is not optimized at this stage. The transient results show the reactor power slowly stabilizing towards a 3% power level after the partial loss of secondary heat removal. Several recriticalities are observed due to cooling through the secondary system but the reactor is self-stabilizing and behaves as expected.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Preliminary Primary System Thermal Fluids Analysis of a Horizontal Compact HTGR

The Horizontal Compact High Temperature Gas Reactor (HC-HTGR) is being designed by a multi-disciplinary team of nuclear, mechanical, and structural engineers, under the support of a DOE-NE Advanced Reactor Demonstration Program’s Advanced Reactor Concepts-20 (ARC-20) award. The objective of this ARC-20 project is to deliver a conceptual design for the proposed HC-HTGR in 3 years and support its commercialization as a safe, low-cost HTGR. Argonne National Laboratory (Argonne) is collaborating on the thermal hydraulic design and analysis of the HC-HTGR reactor pressure vessel internals. The scope of this work includes ensuring the reactor is able to maintain maximum core temperatures below designated safety thresholds during normal operation, shutdown, and accident conditions. This report documents the preliminary thermal hydraulic analysis of the HC-HTGR core design performed with a 1D fluid-3D solid coupled model built using the System Analysis Module (SAM). This assembly level model was utilized to inform the core assembly design, predict the temperature distribution in the peak power assembly including the peak fuel temperature, coolant channel outlet temperatures, and graphite temperature gradients. A key result of this analysis was the determination that the peak fuel temperature remains below the safety threshold of 1250°C. Additionally, this model was used to assess the assembly coolant channel and bypass flow mass flow rate distribution. Following the assembly level analysis, attention turned to the development of a full core reduced order model to be used to predict the core wide coolant flow distribution and to model certain operational and accidental transients.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Measurement of the mass-changing, charge-changing, and production cross sections of C 11 , B 11 , and B 10 nuclei in C 12 + p interactions at 13.5 GeV / c per nucleon

A good knowledge of nuclear fragmentation cross sections is important to interpret the fluxes of secondary cosmic rays from the Galaxy. Here we report new measurements of nuclear fragmentation with the NA61/SHINE experiment at the CERN SPS. The specific focus is on cross sections important for the production of boron in the Galaxy from the interactions of C 12 nuclei with hydrogen in the interstellar medium, including the contribution from the decay of the short-lived C 11 fragments. The data were taken with the secondary C 12 beam at beam momentum of 13.5 GeV / c per nucleon and two fixed targets, polyethylene ( CH 2 ) and graphite (C), from which we derive the cross sections of carbon on hydrogen. We present the measurement of the fragmentation cross sections of C 11 , B 11 , and B 10 as well as the mass- and charge-changing cross sections.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1976

Research on plasma core reactors

Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diameter by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17 cm thick by 89 cm diameter beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000 cu cm aluminum canister in the central region was fueled with UF6 gas and fission density distributions determined. These results are to be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

Jarvis, G. A.↗
NASA NTRS1979

Initial conceptual design study of self-critical nuclear pumped laser systems

An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

Rodgers, R. J.↗
DOE OSTI2021

ECAR-1406 AS-RUN NEUTRONIC ANALYSIS OF THE AGC-1 EXPERIMENT IRRADIATED IN THE ATR SOUTH FLUX TRAP

This Engineering Calculations Analysis Report (ECAR) documents the results of the Advanced Test Reactor (ATR) detailed physics analyses performed to calculate the displacements per atom (DPA) and the fast neutron fluence (E > 0.1 MeV) of the Advanced Graphite Creep (AGC) experiment, AGC-1, irradiated in the ATR South Flux Trap (SFT) (see Figure 1) during ATR Cycle 145A, Cycle 145B, Cycle 146A, Cycle 146B, Cycle 147A, Cycle 148A and Cycle 148B. The results for these evaluations and analysis are reported herein.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

NDMAS Portal Updates

Overview of current ART-GCR Data, including fuel fabrication, irradiation monitoring, fuel and graphite, post-irradiation examination, graphite characterization, high temperature metals mechanical tests, methods validation data (including: HTTR, NSTF, and HTTF), addition of AGR fuel data, other GCR and related data, updated high temperature metals, ongoing and upcoming work, users guide, and the link to those interested in access to NDMAS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1975

Beryllium reflected cavity reactor for UF6 critical experiments

Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diam by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials are available. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

Jarvis, G. A.↗
DOE OSTI2020

Recent Testing of ENDF/B-VIII.0 at ORNL [Slides]

This presentation provides an overview of comparisons of ENDF/B-VII.1 and ENDF/B-VIII.0 and uses VALID suite to test >600 cases with ENDF/B-VII.1 and ENDF/B-VIII.0 in KENO V.a/KENO-VI, although a limited number of cases have been tested with TSUNAMI. It provides a results summary of KENO V.a/KENO-VI with a pooled results summary. The presentation also discusses changes for LCT experiments, MCT experiments, MST experiments, and PST experiments. A few words about graphite are also provided and the presentation states that graphite performance investigated across ENDF/B-VII.0, ENDF/B-VII.1, and ENDF/B-VIII.0 for MSR and HTGR models. A brief summary of the results is presented to make sure the Validation Committee is aware of the work. Also presented is an Eigenvalue comparison with a graphite moderated system, and comparison between an HTR-10 pebble model ENDF/B-VII.1 vs. ENDF/B-VIII.0, which provides different graphite data. In conclusion, ORNL has recently completed a comparison of ENDF/B-VIII.0 and ENDF/B-VII.1 using the VALID library. Work is also being performed for advanced reactor applications. Other efforts on-going and they are also trying to investigate more intermediate spectrum cases.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Initial Developments in Modeling Graphite Behavior

Graphite is used in High Temperature Reactor designs as a moderator, reflector and core material responsible for protecting the fuel and maintaining structural stability. From a safety perspective, it is imperative to be able to predict how graphite will behave under reactor operating conditions which may compromise the core’s structural integrity or fuel’s safety performance. The following report summarizes recent modeling work undertaken at Idaho National Laboratory which focuses on graphite degradation behavior. One phenomenon which is explored in this report and can cause significant degradation to graphite properties is oxidation. Oxidation behavior in graphite is highly temperature dependent. At low temperatures, oxygen can fully penetrate a component and cause a homogeneous damage profile. At high temperatures, only the graphite near the surface of a component is affected. Understanding and modeling this temperature dependence is essential to predicting graphite behavior. In this report, three graphite models are discussed. The first model is used to investigate strength loss after low temperature oxidation. It does this by generating an approximate graphite microstructure then determining the required load to cause crack propagation. It has been shown that the model reproduces the quasi brittle stress versus strain relationship observed in graphite. Strength loss results from the model are shown to match experimental values. The second model investigates strength loss under high temperature oxidation conditions. This model is more applicable to full scale graphite components where the oxidation damage is often inhomogeneous throughout a component. The third model computes stresses in graphite under reactor conditions. Specifically, the model incorporates the effect of creep, irradiation dose, thermal strains, and oxidation in order to predict internal stresses in a component. All of these models are implemented in the Multiphysics Object-Oriented Simulation Environment (MOOSE), an open source, parallel finite element framework.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Multiphysics Simulations of MSRE with NEAMS Thermal Hydraulics Tools

This report documents the benchmarks being developed and simulations performed using tools and codes developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, utilizing MSRE experimental data. In FY23, three main work scopes were investigated under the NEAMS MSR work package at ANL. The first scope investigated the Griffin-SAM coupling model for simulating the pump startup transient experiment of MSRE. The analyses start with a simple model (single-channel, single-lattice), gradually adding more details (multi-channel, full-core) into the model. The results show that the reactivity loss curve is very sensitive to the axial boundary conditions and the radial core discretization. The simple model can predict a similar reactivity trend as that of the more sophisticated model, which is likely due to error cancellation. Accurately modeling the axial boundary condition may further improve the reactivity trend but would require significant efforts to generate the mesh of the MSRE inlet and upper plenum. The core channel radial discretization for the Griffin-SAM coupled model also depends on the flow distribution. Given the complex geometry in the inlet plenum, the flow distribution needed to be calculated from CFD analysis, which was performed using the NekRS code. This analysis employed a MSRE CAD model developed by Copenhagen Atomics. The CAD model was disassembled to keep the inlet plenum region only, which was subsequently cleaned and modified so that the mesh generated is under the memory limit. The results are merged to a few radial regions to show that the flow rate is highest in the central region. This would be useful for future improvement of the Griffin-SAM coupling model of the MSRE core. The last task investigated is tritium transport modeling using the standalone SAM code. This task aimed to initiate the effort to demonstrate and validate the tritium transport model implemented in SAM. The preliminary investigation employed an MSRE model consisting of the primary loop. Three tritium transport pathways were examined including the retention in the graphite, the permeation through the HX tube wall, and the removal from the off-gas system. The results compare well with the MSRE data, but improvements are still needed on the initial conditions (i.e., the present state may not have reached equilibrium), the boundary conditions, the off-gas system modeling, and a better numerical strategy to reach the equilibrium state.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS2014

Low-Cost Radiator for Fission Power Thermal Control

NASA Glenn Research Center (GRC) is developing fission power system technology for future Lunar surface power applications. The systems are envisioned in the 10 to 100kW(sub e) range and have an anticipated design life of 8 to 15 years with no maintenance. NASA GRC is currently setting up a 55 kW(sub e) non-nuclear system ground test in thermal-vacuum to validate technologies required to transfer reactor heat, convert the heat into electricity, reject waste heat, process the electrical output, and demonstrate overall system performance. Reducing the radiator mass, size, and cost is essential to the success of the program. To meet these goals, Advanced Cooling Technologies, Inc. (ACT) and Vanguard Space Technologies, Inc. (VST) are developing a single facesheet radiator with heat pipes directly bonded to the facesheet. The facesheet material is a graphite fiber reinforced composite (GFRC) and the heat pipes are titanium/water. By directly bonding a single facesheet to the heat pipes, several heavy and expensive components can be eliminated from the traditional radiator design such as, POC(TradeMark) foam saddles, aluminum honeycomb, and a second facesheet. A two-heat pipe radiator prototype, based on the single facesheet direct-bond concept, was fabricated and tested to verify the ability of the direct-bond joint to withstand coefficient of thermal expansion (CTE) induced stresses during thermal cycling. The thermal gradients along the bonds were measured before and after thermal cycle tests to determine if the performance degraded. Overall, the results indicated that the initial uniformity of the adhesive was poor along one of the heat pipes. However, both direct bond joints showed no measureable amount of degradation after being thermally cycled at both moderate and aggressive conditions.

Bonding↗
DOE OSTI2025

Hot Hydrogen Exposure of U x Zr 1-x C y Nuclear Fuel: The Effect of Additional Carbon

Nuclear thermal propulsion (NTP) using hydrogen as propellant in a solid-state fission reactor to reach temperatures of up to 3000K can achieve significantly higher specific impulse than chemical propulsion. A promising fuel that is again considered for NTP is U x Zr 1-x C y . To reduce the overall mass loss as well as the uranium fuel loss, a range of experiments using additional carbon in hot hydrogen (2600 K, 6 SLPM flow rate) were performed for up to 5-6 h. Both a CH 4 addition (0.1 and 0.2 vol%) in the hydrogen stream and the use of sacrificial graphite pieces upstream of the samples were tested, with sample compositions from 5 at% UC to 30 at% UC. The results showed a strong reduction of mass loss, up to a factor of 3, as well as a reduction of surface uranium losses in the presence of additional carbon in the atmosphere. Even with additional carbon in the atmosphere, material with 30 at% UC was not stable, but material with 20 at% UC was stable for up to 6h. The sample surfaces showed texturing after processing, possibly by grain growth on the surface. This led to a separation into grains with high uranium content close to the (100)-orientation, and with low uranium content in grains close to the (111)-orientation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Incident beamline design for a modern cold triple axis spectrometer at the High Flux Isotope Reactor

A modern cold triple axis spectrometer is being planned for the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Here, we describe the design of an incident beamline that will put a flux of ~ 10 8 $\frac {n}{cm^{2}s}$ on a sample with an area of 2 cm X 2 cm. It takes current physical constraints at HFIR into account and it can accommodate both single and multiplexed analyzer-detector secondary spectrometers and large superconducting magnets. The proposed incident beamline includes a multi-channel guide with horizontal focusing, a neutron velocity selector, components to facilitate an incident beam polarization option, and a double-focusing pyrolytic graphite monochromator. Here, this work describes the process of optimizing the guide system and monochromator and summarizes the expected performance of the incident beamline for non-polarized operation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Gear Test Assembly - Report on Experimental Testing (FY2020)

The Gear Test Assembly (GTA) has completed Campaign #2 of in-sodium testing at the Mechanisms Engineering Test Loop (METL). Campaign #1 was conducted in 2019 and completed 19,600 cycles, correlating to 9,800 fuel assembly maneuvers. Campaign #1 was terminated following a shaft seizure caused by a failed bearing in the sodium gearbox. Campaign #2 was conducted in the spring of 2020 and completed 2,768 cycles, correlating to 1,384 fuel assembly maneuvers. Campaign #2 was terminated following elevated vibration measurements coming from the sodium gearbox. Following cleaning and disassembly, a bearing was found to have failed in the sodium gearbox. Campaign #1 used tapered roller bearings made of 52100 steel with a heat treatment from the manufacturer that allows for operation to 350°C. Campaign #2 used the same tapered roller bearings, but with no heat treatment. A new dynamic shaft seal design was used in Campaign #2 that led to significant reductions in argon consumption and graphite debris generation. A modified external gearbox system led to no external bearing failures and allowed for higher torque output. Finally, modifications to the Carbonation System led to a more effective carbonation process and all-around reduction in turnaround time.

42 ENGINEERING↗
DOE OSTI2022

A Subcritical Assembly for Training and Education Use at the Oak Ridge National Laboratory [Slides]

Sufficient AGN-201M fuel plate material exists at Y-12 to support the development of a new, inherently safe, subcritical assembly for use in NCS training courses at ORNL. Four experiments with the ORNL subcritical assembly are possible and feasible. Future work includes the final design of the core and reflector and split table component design specifications. The ORNL Material Demonstration Facility (MDF) may be used to 3D print ORSA stand/table components and graphite reflector. Researchers will consider fuel coatings for contamination control, rather than an Al can. A visit to Y-12 will be made to determine the location of AGN fuel plates and remaining fuel transportation considerations. Final facility location will be considered

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗