Digital codes for design and evaluation of convectively cooled rocket nozzle with application to nuclear-type rocket
Computer programs for design of convectively cooled rocket nozzle with application to nuclear type rocket
COMPUTER PROGRAM↗
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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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An evolution strategy for lunar nuclear surface power
The production and transmission of electric power for a permanently inhabited lunar base poses a significant challenge which can best be met through an evolution strategy. Nuclear systems offer the best opportunity for evolution in terms of both life and performance. Applicable nuclear power technology options include isotope systems (either radioisotope thermoelectric generators or dynamic isotope power systems) and reactor systems with either static (thermoelectric or thermionic) or dynamic (Brayton, Stirling, Rankine) conversion. A power system integration approach that takes evolution into account would benefit by reduced development and operations cost, progressive flight experience, and simplified logistics, and would permit unrestrained base expansion. For the purposes of defining a nuclear power system evolution strategy, the lunar base development shall consist of four phases: precursor, emplacement, consolidation, and operations.
Mason, Lee S.↗
How Bayesian methods can improve R -matrix analyses of data: The example of the d t reaction
The 3 H(d, n) 4 He reaction is of significant interest in nuclear astrophysics and nuclear applications. It is an important, early step in big-bang nucleosynthesis and a key process in nuclear fusion reactors. We use one- and two-level R-matrix approximations to analyze data on the cross section for this reaction at center-of-mass energies below 215 keV. We critically examine the data sets using a Bayesian statistical model that allows for both common-mode and additional point-to-point un- certainties. We use Markov Chain Monte Carlo sampling to evaluate this R-matrix-plus-statistical model and find two-level R-matrix results that are stable with respect to variations in the channel radii. The S factor at 40 keV evaluates to 25.36(19) MeV b (68% credibility interval). We discuss our Bayesian analysis in detail and provide guidance for future applications of Bayesian methods to R-matrix analyses. We also discuss possible paths to further reduction of the S-factor uncertainty.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Additive manufacturing: a challenge to nuclear nonproliferation
Additive manufacturing (AM) is a rapidly developing technology that allows industries to automate and simplify the production of highly complicated items. Recently, there are concerns that AM techniques have the potential to be utilized in the fields of nuclear weapons and nuclear enrichment technology. Presently, there are very few international or domestic export controls that apply to AM's role in the nuclear industry, constituting an unmanaged proliferation pathway. Pre-existing export controls focus on general concepts and processes and do not consider specific nuances of various techniques that are categorized as additive manufacturing. To introduce legislation and controls that will be effective in monitoring proliferation pathways, one must investigate and characterize AM techniques and their nuclear applications. This paper categorizes and ranks 33 AM techniques based on their potential impact on the nuclear fuel cycle and the development of nuclear weapons. Through this method of characterization and categorization, export controls would address specific AM nuclear proliferation risks without disrupting the entire industry and fuel cycle. Additionally, legislation employing this method would identify loopholes in export controls using a holistic approach to managing and monitoring proliferation pathways.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Updates to the n+ 63,65 Cu Evaluations in the Resolved Resonance Region [Slides]
This presentation discusses the motivation and background of the n+ 63,65 Cu Evaluations in the Resolved Resonance Region which is to study the interaction of neutrons with copper as it is important in nuclear applications since critical assembly configurations include metallic copper as reflector. In support to the U.S. Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP), measurements and related evaluations of 63,65 Cu isotopes were selected to improve the agreement with the benchmarks and to assess the importance of the angular distribution data for reactor calculations. Previous and current evaluation work is supported by an experimental campaign initiated before 2010, the 63,65 Cu R-matrix analysis generated resonance parameters up to 300 keV. However, due to outstanding issues in the benchmark performance, ENDF/B-VIII.0 library released a truncated set of resonance parameters up to 100 keV. The goal of this work is to generate an updated set of resonance parameters in the 100-300 keV range to improve the benchmark performance of 63,65 Cu isotopes. In conclusion, R-matrix analysis to update 63,65 Cu evaluations was performed to simultaneously improve benchmark performance and extend the RRR to 300 keV. The benchmark calculations suggest the increased capture cross sections are beneficial, however, further investigation of the measured capture data is needed to understand the large normalization scaling factor needed to improve the reactivity. Also, the copper-reflected benchmarks indicate the need to further investigate angular distributions and extension of RRR to 300 keV is aided well by level statistics considerations. Work to refine the fit of individual resonances is ongoing.
07 ISOTOPE AND RADIATION SOURCES↗
A nuclear driven metallic vapor MHD coupled with MPD thrusters
Nuclear energy as a source of power for space missions, represents an enabling technology for advanced and ambitious space applications. Nuclear fuel in a gaseous or liquid form has been configured as a promising and practical candidate in this regard. The present study investigates and performs a feasibility analysis of an innovative concept for space power generation and propulsion. The system embodies a conceptual nuclear reactor with an MHD generator and coupled to MPD thrusters. The reactor utilizes liquid uranium in droplet form as fuel and superheated metallic vapor as the working fluid. This ultrahigh temperature vapor core reactor brings forward varied and challenging technical issues, and it has been addressed to in this paper. A parametric study of the conceived system has been performed in a qualitative and quantitative manner. Preliminary results show enough promise for further indepth analysis of this novel system.
Anghaie, Samim↗
Radiation Hardened Electronics for Reactor Environments
Radiation-hardened (rad-hard) electronics have been the shortcoming in nuclear sensing and instrumentation. Placing advanced sensors and associated electronics closer to a nuclear reactor core will improve reactor control and operation through increased signal accuracy, precision, and fidelity, resulting in safer, more efficient energy production. However, the extreme temperature and bombarding radiation environment of nuclear applications make research and development into electronics and electronic materials technologies essential to enabling improved safety, monitoring, and control of the existing nuclear reactor fleet and the next generation of reactors, including microreactors. To help the US Department of Energy (DOE) define a course for the Office of Nuclear Energy (NE)-funded rad-hard electronics research, Oak Ridge National Laboratory staff present this summary of the state-of-the-art (SOA) technology for the design and implementation of rad-hard electronics and systems, including research efforts and commercial offerings. A collection of technological gaps associated with reactor applications are compiled herein. Based on the radiation degradation mechanisms discussed in this report, along with the state-of-the-art summary survey and a compiled list of gaps, a high-level research plan of future NE funding direction was developed and is provided. This report outlines these activities and provides a summary of SOA electronics technology for rad-hard instrumentation in reactors, a list of the major technology gaps observed, and a draft plan for addressing the present and future needs of reactor instrumentation.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Nuclear Data Adjustment with Whisper for Criticality Safety Applications [Slides]
Nuclear data adjustment is used by Whisper to compute residual nuclear data uncertainties within a full suite of benchmarks; in order to do so, a complete set of nuclear data covariances, benchmarks and k-effective sensitivity profiles (for both benchmarks and applications) is required. For applications that are similar to the full benchmark suite, the residual uncertainties add a small margin to the overall USL calculation. For applications that are dissimilar to the full benchmark suite, the residual uncertainties add a reasonable margin to the overall USL calculation (more conservative). This methodology helps to overcome any issues in the released covariance libraries where nuclear data uncertainties appear to be too large.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Model uncertainty in accelerator application simulations
Monte-Carlo nuclear reaction and transport codes are widely used to devise accelerator-based nuclear physics experiments; at the same time, many experiments are performed to validate the Monte-Carlo codes, which can be used for the design of full-scale nuclear power applications or the design of new benchmark experiments. Dedicated model benchmark studies investigate a broad range of nuclear reactions and quantities. Examples of these include isotope formation or secondary particle fluxes that result from the interactions of GeV-range hadrons with monoisotopic targets, which can be used to assess the respective systematic uncertainty of models. Such benchmark studies, as well as many nuclear application experiments and simulations carried out by various groups over the last few decades, enable us to draw methodological lessons. In this work, model uncertainty determined based on available experimental data allow us to identify the effects of practitioner expertise as well as the design of codes (user access to micro-scale parameters) on the range of uncertainties. We found that in cases when simulations are performed by code developers or users that are very experienced in performing simulations, the model to experiment quantity ratios generally agree with the limits determined by dedicated benchmark studies. In other cases, the ratios generally tend to be either smaller (underestimation of model error) or larger (overestimation of model error). A plausible explanation of the aforementioned effects is suggested.
72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
Experiment design for the neutron irradiation of $\mathrm{PM-HIP}$ alloys for nuclear reactors
Here, this article describes the design of an Advanced Test Reactor (ATR) drop-in neutron irradiation experiment aiming to directly compare the performance of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP) against conventional casting or forging. There is growing interest in PM-HIP alloys for nuclear applications because of their microstructural uniformity, superior mechanical properties, and reduced dependence on welding and machining, compared to cast/forged alloys. Nuclear code-qualification of PM-HIP alloys requires neutron irradiation testing to demonstrate performance under relevant conditions. In this experiment, six nuclear structural alloys were irradiated: Ni-based alloys 625 and 690, Grade 91 ferritic steel, SA508 pressure vessel steel, and 304L and 316L austenitic stainless steels. The experiment is assembled into seven capsules in four test trains and irradiated in three ATR inboard A positions. Both the PM-HIP and cast/forged versions of each alloy were irradiated under nearly identical conditions for comparative purposes, to target doses of 1 ± 0.2 and 3 ± 0.2 dpa at temperatures of 300 ± 50 °C and 400 ± 50 °C. A thorough description of the experiment design and thermal, structural and neutronic analyses performed to ensure the targeted irradiation conditions are met is provided. Specimens were configured as small disks, compact tension specimens and tensile bars to facilitate post-irradiation examination (PIE) that will include mechanical testing, microstructure characterization, and fracture toughness testing. Given the considerations for ASTM standardized mechanical testing, comparative fluence and temperature across specimen pairs, and comprehensive PIE planning herein, this work serves as a template for future nuclear materials qualification experiment designs.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
A 1050 K Stirling space engine design
As part of the NASA CSTI High Capacity Power Program on Conversion Systems for Nuclear Applications, Sunpower, Inc. completed for NASA Lewis a reference design of a single-cylinder free-piston Stirling engine that is optimized for the lifetimes and temperatures appropriate for space applications. The NASA effort is part of the overall SP-100 program which is a combined DOD/DOE/NASA project to develop nuclear power for space. Stirling engines have been identified as a growth option for SP-100 offering increased power output and lower system mass and radiator area. Superalloy materials are used in the 1050 K hot end of the engine; the engine temperature ratio is 2.0. The engine design features simplified heat exchangers with heat input by sodium heat pipes, hydrodynamic gas bearings, a permanent magnet linear alternator, and a dynamic balance system. The design shows an efficiency (including the alternator) of 29 percent and a specific mass of 5.7 kg/kW. This design also represents a significant step toward the 1300 K refractory Stirling engine which is another growth option of SP-100.
Penswick, L. Barry↗
Report on the Assessment of Ion Irradiations on High Entropy Alloys
Idaho National Laboratory (INL) initiated a joint international effort with the Czech Republic to explore the feasibility of manufacturing High Entropy Alloys for high-temperature nuclear applications using Additive Manufacturing. This effort was funded at INL by the United States Department of Energy’s Office of Nuclear Energy under the Advanced Reactor Technologies and Advanced Materials and Manufacturing Technologies (AMMT) Program. The High Entropy Alloys were specifically designed for the corrosive and irradiation environments as experienced in gas fast reactors, molten salt reactors, and fusion power. These alloys were manufactured by multiple processes to determine the impact of manufacturing processes on the performance of the alloys in corrosive and irradiation environments. Preliminary results show that MoNbTiV and MoNbTi alloys exhibit no degradation after multiple corrosion tests. These MoNbTiV and MoNbTi alloys were also evaluated after irradiation experiments were conducted at the Michigan Ion Beam Laboratory at the University of Michigan, which is the focus of this milestone report. For these irradiation experiments, the MoNbTiV and MoNbTi alloys manufactured by arc melting and Electric Field Assisted Sintering (EFAS) were tested at 500 C at a dose of 30 dpa to evaluate the microstructural stability and decomposition.
36 - MATERIALS SCIENCE↗
The brayton cycle for space power
This paper reviews the basic features of the Brayton cycle and discusses the selection of such cycle parameters as working fluid, pressure, and temperature. System performance, weight, and reliability characteristics are then considered in order to determine the suitability of this type of system for the intended space applications. It is shown that the Brayton cycle can be considered suitable for low-power solar and intermediate-power nuclear applications where low specific weight is not a critical requirement. To achieve the low specific weights required for electrical propulsion missions, turbine-inlet temperatures in the 3000 R region at power levels in the 500–1000 kw range appear to be necessary.
ELECTRIC PROPULSION↗
Burnable absorbers in nuclear reactors – A review
Burnable absorbers can benefit nuclear reactors of virtually any design by providing reactivity control for extended fuel cycles, tritium production, burning of long-lived radionuclides, and reactor safety. When selecting the ideal burnable absorber type and its design, one must consider the resulting impact on the reactor’s fuel cycle design and cost, reactivity, thermal hydraulics, manufacturing, and radiation response. These selection criteria, as well as neutronic and thermophysical material property requirements, may be vastly different depending on whether the burnable absorber is intended for use in a commercial water-cooled reactor, a research reactor, or a next-generation advanced reactor system. A recent integration, synthesis, description of past and present technologies, and identification of existing gaps and areas of future research is lacking on these important topics. Here, this paper includes a fundamental description of the use of burnable absorbers and their impacts on reactivity, absorber depletion, self-shielding, basic thermophysical properties, and the use of burnable absorbers in next-generation nuclear applications..
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
FY 2025 Multidimensional Data Correlation Platform: Unified Software Architecture for Advanced Materials and Manufacturing Technologies Data Management and Processing
The Advanced Materials and Manufacturing Technologies (AMMT) program continues to advance a data-driven approach to demonstrate the utility of additive manufacturing for fabricating components for nuclear applications. A key scientific goal is to leverage data to better understand manufacturing outcomes and thereby improve the performance, reliability, and lifespan of nuclear components. Ultimately, this effort supports the development of standards for certification and qualification of additively manufactured components, enabling broader industry adoption. In support of this objective, the AMMT program is building and deploying a data management platform to record, index, analyze, and make available the manufacturing data generated across the AMMT program. In FY 2023, the team conceptualized the architecture of the platform and, in FY 2024, deployed the first functional version at the Oak Ridge National Laboratory (ORNL) Manufacturing Demonstration Facility (MDF). In FY 2025, the platform was officially opened to all AMMT members. To enable this expansion, core modifications and enhancements were developed, including improvements to the user interface and workflows for data entry and retrieval. Most notably, robust security and access control mechanisms were implemented to protect data and manage information sharing. This effort featured a logging system, protected views, and controlled access mechanisms. This report documents these enhancements and the transition of the platform into program-wide use.
36 MATERIALS SCIENCE↗
Behavior of the Fused Quartz Suspended Core Acoustic Waveguide Sensor in Gamma Radiation
Previously, a fused-quartz acoustic waveguide named the suspended core waveguide with tight-field confinement in a small diameter was developed and its sensing potential in harsh environment was demonstrated with the fabricated periodic acoustic fiber Bragg grating. In this paper, the acoustic sensor was exposed to around 70 Gy/h gamma radiation at room temperature for 3550 h at Oak Ridge National Laboratory. For comparisons, a thermocouple and two optical fiber Bragg gratings were tested under the same conditions. The central frequency reading of the acoustic sensor was found to encounter a relatively fast decrease of 0.5 kHz in the first 800 hours. The reading then became stable, centering at 478.5 kHz with a fluctuation of ±0.2 kHz, and responded to small environmental temperature variations less than 1.2 °C. The major effect to the fluctuation was concluded to be radiation-induced material compaction and expansion. For optical fiber Bragg gratings, the same one-directional fast change of readings in the first 500 hours were observed as well. Although the optical gratings were able to track temperature changes, they also showed continuous drifts. The survivability and consistency of the acoustic sensor under long-term gamma radiation could lead to new sensing methods in nuclear applications.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Local chemical ordering of a neutron-irradiated CrFeMnNi compositionally complex alloy
While ion-irradiation studies are a critical first step in studying compositionally complex alloys (CCAs) for nuclear applications, they do not capture all the microstructural changes occurring under the low irradiation dose rates and different particles’ scattering patterns experienced in a nuclear reactor setting. To explore these phenomena in reactor-relevant conditions for the first time in CCA, the single-phase solid-solution Cr 10 Fe 30 Mn 30 Ni 30 was neutron irradiated up to 6.61 displacements per atom at 395 and 579 °C. Irradiation-enhanced local chemical ordering (LCO) well beyond the range of short range ordering was observed, and is predicted to be the precursor to the precipitation of a coherent Ni-Mn L1 0 phase and a Cr-rich α’ phase, though TEM analysis did not indicate the presence of either in any irradiation condition. The line density of faulted dislocation loops decreased from 6.47 to 1.69 ∙ 10 15 m -2 from 3.43 to 6.61 dpa at 579 °C despite no appreciable faulted loop content in the unirradiated material. LCO is expected to increase the complexity of the energy landscape within this alloy, restricting interstitial point defect mobility and creating local regions of greater stacking fault energy. These contribute to the negative correlation between irradiation dose and faulted dislocation loop density in this study, as well as the lack of void swelling observed.
36 MATERIALS SCIENCE↗
UO 2 + 5 vol% ZrB 2 nano composite nuclear fuels with full boron retention and enhanced oxidation resistance
The boron isotope ( 10 B) can be used as a neutron absorber in UO 2 to control the reactivity of nuclear fuel pellets, however, the boron source can react with oxygen source in UO 2 to form B 2 O 3 that vaporize readily at temperatures above 1200 °C. Unfortunately, the sintering of UO 2 fuel requires hours holding at high temperature (>1700 °C), resulting in an inevitable B loss during sintering and unpredictable B concentration in final product. It is challenging to incorporate boron through a conventional sintering method. In this work, we demonstrated that spark plasma sintering (SPS), a field assisted sintering technology, can effectively densify UO 2 + 5 vol% ZrB 2 composite fuel pellets by rapid consolidation at 1600 °C for a short duration of 5 min under an applied pressure of 40 MPa. Further, thermogravimetric analysis (TGA) measurements confirm that ZrB 2 is fully retained inside the composite fuel pellets. Inside the composite fuel pellets, nano sized ZrB 2 particles are uniformly distributed along the grain boundaries of the UO 2 matrix. The ZrB 2 particle transforms to a glassy B 2 O 3 phase covering the sample surface and grain boundaries of UO 2 matrix after a simple post-sintering annealing at 1000 °C in flowing Argon gas for 4 h. The formed glassy B 2 O 3 slows down the diffusion of oxygen ions and postpones the onset temperature for oxidation of UO 2 from 400 °C to 550 °C. This study demonstrates the capability of SPS, an advanced fuel manufacturing technique, to achieve a full retention of ZrB 2 in UO 2 oxide fuel and increase oxidation resistance through a simple post-sintering annealing. The reported work holds great engineering potential for development of advanced oxide fuel for nuclear application.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗