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DOE OSTI2024

Thermodynamics of Tritium Trapping by Point Defects in Intermetallic Al 12 (TM) 2.35 Aluminide Coating Phase

Density functional theory simulations have been carried out to investigate the potential for tritium trapping by metal vacancies in intermetallic Al 12 (TM) 2.35 phase (TM = Fe, Cr, and Ni) as function of temperature and tritium partial pressure. It was found that tritium could be favorably trapped by Fe and Ni vacancies and not favorably trapped by Al and Cr vacancies. However, due to the presence of partially occupied Al sites in bulk Al 12 (TM) 2.35 , leading to the approximate number of ~255 Al atoms in the unit cell, 86 sites were found energetically favorable to the creation of an Al vacancy. While adding a tritium atom in an Al vacancy is not energetically favorable, the tritiated defect still has a negative Gibbs free energy because the energy gain for creating an Al vacancy overcome the energy cost of adding the tritium species. Based on the calculated Gibbs free energy, the first tritiation of a metal vacancy, at conditions relevant to in-reactor operations, should be more favorable for Al, followed Fe, Ni, and Cr vacancies. By comparing the behavior of tritium in Al 12 (TM) 2.35 with previously studied Fe-Al coating phases (i.e., FeNiAl 5 , Fe 4 Al 13 , and Fe 2 Al 5.6 ), we found that there is a correlation between interstitial tritium solubility and the potential for vacancy trapping. The current trend suggests that if the insertion of an interstitial tritium cost more than 0.3 eV, then trapping by metal vacancies should be preferred. By combining the simulations results obtained to date, we noticed different trapping mechanisms of tritium in the Al coating. Tritium is mostly trapped by Fe and Ni vacancies in the outer Fe-Al coating phase Al 12 (TM) 2.35 while tritium should be preferentially trapped by Al and Fe vacancies for the inner Fe-Al coating phases (FeNiAl 5 , Fe 4 Al 13 , Fe 2 Al 5.6 ). Altogether, these studies show that tritium interacts differently with the various Fe-Al aluminide phases, they also suggest that tritium trapping and retention could be more efficient if metal defects are present and if the solubility of interstitial tritium in the different phases is low.

36 MATERIALS SCIENCE↗

DOE OSTI2025

SiC cladding R&D execution plan

To advance the development of accident-tolerant fuel cladding for light-water reactors, the Advanced Fuels Campaign actively addresses the R&D gap between the specialized technological progress in SiC-based composite materials, led by industrial partners, and the fundamental scientific understanding of material behavior under reactor operating conditions. The laboratory activities are systematically organized based on the Development Strategy document, which outlines critical barriers to the use of SiC composites in nuclear energy applications for the coming decade and beyond. To complement the strategy, this execution plan specifies a detailed year-by-year work package and establishes key milestones for assessing progress in technological advancement. This plan has been formulated in collaboration with principal investigators at national laboratories, stakeholders, and industry partners with the goal of closing technology gaps and readying SiC-based cladding technologies for full-scale commercial testing. The research conducted at the national laboratories is categorized into three primary subjects: (1) addressing challenges associated with the performance of SiC cladding under normal operating environments, (2) identifying the failure limits of SiC cladding under accident conditions, and (3) modeling fuel performance for SiC cladding systems. Comprehensive laboratory activities integrate the development of advanced modeling tools for SiC-based cladding systems with experimental validation via integrated testing of multiphysics phenomena. This document lays out the laboratory plan to support the development of SiC-based technologies, which have the potential for significant operational and safety benefits for the U.S. nuclear industry. This plan aligns with industry developer needs, relevant Nuclear Energy University Programs, and Nuclear Energy Advanced Modeling and Simulation Programs to ensure SiC cladding technologies are matured as efficiently as possible and will be maintained and updated to accommodate the dynamically evolving progress of development, capability needs, and resource availability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Radiological Safety Analysis Computer (RSAC) Program Version 7.4 Users’ Manual

The Radiological Safety Analysis Computer (RSAC) Program Version 7.4 (RSAC-7) is the newest version of the RSAC legacy code. RSAC-7 calculates the consequences of a release of radionuclides to the atmosphere. Users generates a fission product inventory from either reactor operating history or a nuclear criticality event. RSAC-7 models the effects of high-efficiency particulate air filters or other cleanup systems and calculates the decay and ingrowth during transport through processes, facilities, and the environment. Doses are calculated for inhalation, air immersion, ground surface, ingestion, and cloud gamma pathways. RSAC-7 is used as a tool to evaluate accident conditions in emergency response scenarios, radiological sabotage events, and safety basis accident consequences. This users’ manual contains the mathematical models and operating instructions for RSAC-7. Instructions, screens, and examples are provided to guide the user through the functions provided by RSAC-7. This program is designed for users who are familiar with radiological dose assessment methods.

73 - NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2025

Challenge of Fielding Experiments at the Annular Core Research Reactor

Conference is the annual Test, Research &Training Reactors (TRTR) meeting. Typically attended by dozens of research reactor operators, educators, trainers, maintenance & service contractors Also, members of the Nuclear Regulatory Commission (NRC) attend. Presentation highlights ACRRs capability and the process used to submit and approve experimental activities. This presentation is adapted from my previous presentation to the Material Test and Research conference, SAND2025-1775526.

Martin, Lonnie E. [Sandia National Laboratories (S↗

DOE OSTI2019

Toxic consequences of a sodium fire in a FFTF dump heat exchanger (Initial release 02-11-2000)

A calculation of sodium aerosol concentrations at the site boundary. This would result from a worse case DHX module tube leak with the reactor operating, DHX fans running, and module dampers open. Resulting site boundary concentrations are less than the risk guidelines for an extremely unlikely event. This will be used in an upcoming revision of the FFTF FSAR and FHA Sodium Fire Analyses.

DHX↗

DOE OSTI2023

A mechanistic model of a PWR-based nuclear power plant in response to external hazard-induced station blackout accidents

Natural hazard-induced nuclear accidents, such as the Fukushima Daiichi Accident that occurred in Japan in 2011, have significantly increased reactor safety studies in understanding nuclear power plant (NPP) responses to external hazard events such as earthquakes and floods. Natural hazards could cause the loss of offsite power in nuclear power plants, potentially leading to a Station Blackout (SBO) accident that significantly contributes to the overall risk of nuclear power plant accidents. Despite the fact that extensive research has been conducted on the station blackout accident for nuclear power plant, further understanding of these events is needed, particularly in the context of the dynamic nature of external hazards such as external flooding. This paper estimates the progression of station blackout events for a generic pressurized water reactor (PWR) in response to external flooding events. The original RELAP5-3D model of the Westinghouse four-loop design pressurized water reactor was adopted and modified to simulate the external flood-induced station blackout accident, including the short-term and long-term station blackout scenarios. A sensitivity analysis of long-term station blackout, examining reactor operation times and analyzing key parameters over time, was also conducted in this work. The results of the analyses, especially the critical timing parameters of key event sequences, provide useful insights about the time during the external flooding event, which is important for plant operators to make timely decisions to prevent potential core damage. This paper represents significant progress toward developing an integrated risk assessment framework for further identifying and assessing the effects of the critical sources of uncertainties of nuclear power plant under external hazard-induced events.

Liu, Tao↗

DOE OSTI2024

Phase equilibria of advanced technology uranium silicide-based nuclear fuel

The phases in uranium-silicide binary system were evaluated in regards to their stabilities, phase boundaries, crystal structures, and phase transitions. The results from this study were used in combination with a well assessed literature to optimize the U-Si phase diagram using the CALPHAD method. A thermodynamic database was developed, which could be used to guide nuclear fuel fabrication, could be incorporated into other nuclear fuel thermodynamic databases, or could be used to generate data required by fuel performance codes to model fuel behavior in normal or off-normal reactor operations. The U 3 Si 2 and U 3 Si 5 phases were modeled using the Compound Energy Formalism model with 3 sublattices to account for the variation in composition. The crystal structure used for the USi phase was the tetragonal with an I4/mmm space. Above 450°C, the U 3 Si 5 phase was modeled. The composition of the USi 2 phase was adjusted to USi 1.84 . The calculated invariant reactions and the enthalpy of formation for the stoichiometric phases were in agreement with experimental data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

High y + Shear-Stress Turbulence Implementation for High Flux Isotope Reactor Narrow Channel Flows

The research objective of this work was to improve the engineering predictions of the turbulence characteristics of flows in curved narrow channels. Such channel flows are commonly encountered in nuclear research and test reactors, with one of them being the high-flux isotope reactor (HFIR). Research reactors bear high heat fluxes, and the proper computing of turbulence is paramount for safe and reliable reactor operation. The study builds on the results of a previous direct numerical simulation of turbulence to inform a well-known Reynolds-averaged Navier–Stokes shear-stress turbulence model and improves its accuracy in simulating parallel channel flows. A new formulation of the loss term in the dissipation conservation equation is suggested. Combined with high wall distance computational grids, the new implementation provides a fast-running flow solution, suitable for engineering purposes. Model generalization for parallel channel flows, in a broader range of frictional Reynolds numbers, is suggested by introducing a new form of the model constants.

CFD↗

DOE OSTI2022

Radiolytic Gas Production from Aluminum Coupons (Alloy 1100 and 6061) in Helium Environments—Assessing the Extended Storage of Aluminum Clad Spent Nuclear Fuel

Corrosion of aluminium alloy clad nuclear fuel, during reactor operation and under subsequent wet storage conditions, promotes the formation of aluminium hydroxide and oxyhydroxide layers. These hydrated mineral phases and the chemisorbed and physisorbed waters on their surfaces are susceptible to radiation-induced processes that yield molecular hydrogen gas (H 2 ), which has the potential to complicate the long-term storage and disposal of aluminium clad nuclear fuel through flammable and explosive gas mixture formation, alloy embrittlement, and pressurization. Here, we present a systematic study of the radiolytic formation of H 2 from aluminium alloy 1100 (AA1100) and 6061 (AA6061) coupons in “dry” (~0% relative humidity) and “wet” (50% relative humidity) helium environments. Cobalt-60 gamma irradiation of both aluminium alloy types promoted the formation of H 2 , which increased linearly up to ~2 MGy, and afforded G-values of 1.1 ± 0.1 and 2.9 ± 0.1 for “dry” and “wet” AA1100, and 2.7 ± 0.1 and 1.7 ± 0.1 for “dry” and “wet” AA6061. The negative correlation of H 2 production with relative humidity for AA6061 is in stark contrast to AA1100 and is attributed to differences in the extent of corrosion and varying amounts of adsorbed water in the two alloys, as characterized using optical profilometry, scanning electron microscopy, Raman spectroscopy, and X-ray diffraction techniques.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

FCIC FY18 LT-baseline Feedstock Images

Set of images taken during pilot plant dilute-acid hydrolysis preatreatment reactor operation. Images were taken with a GoPro Hero 5 with rectilinear lens facing downwards on a weigh-belt instrument. Image period was every 30 seconds during operation. This image set is connected to a journal article titled "Real-Time Biomass Feedstock Particle Quality Detection Using Image Analysis and Machine Vision", http://dx.doi.org/10.1007/s13399-020-00904-w.

09 BIOMASS FUELS↗

NASA NTRS1971

Long-duration firings of a Mariner Mars 1969 catalytic reactor

Two long-duration tests were conducted with a surplus Mariner Mars 1969 monopropellant hydrazine reactor in an attempt to induce the washout phenomenon. The Mariner Mars 1969 reactor was chosen because it has a long development history and thus is well characterized. No washout occurred during either of the two 1000-s tests, although slow transients were observed in the reactor operation during what were nominally steady-state conditions. The 2000-s of operating time represents nearly an order of magnitude increase over the rated life of the engine.

Price, T. W.↗

NASA NTRS1972

Neutronics analysis of an open-cycle high-impulse gas core reactor concept

A procedure was developed to calculate the critical fuel mass, including the effects of propellant pressure, for coaxial-flow gas-core reactors operating at 196,600 newtons thrust and 4400 seconds specific impulse. Data were generated for a range of cavity diameter, reflector-moderator thickness, and quantity of structural material. Also presented are such core characteristics as upper limits on cavity pressure, spectral hardening in very-high-temperature hydrogen, and reactivity coefficients.

Whitmarsh, C. L., Jr.↗

NASA NTRS1971

Thermionic reactor power conditioner design for nuclear electric propulsion.

Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

Jacobsen, A. S.↗

NASA NTRS1973

Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation.

Whitmarsh, C. L., Jr.↗

NASA NTRS1974

Advanced-power-reactor design concepts and performance characteristics

Five reactor cooling concepts which allow continued reactor operation following a single rupture of the coolant system are presented for application with the APR. These concepts incorporate convective cooling, double containment, or heat pipes to ensure operation after a coolant line rupture. Based on an evaluation of several control system concepts, a molybdenum clad, beryllium oxide sliding reflector located outside the pressure vessel is recommended.

Davison, H. W.↗

NASA NTRS1974

Physics and potentials of fissioning plasmas for space power and propulsion

A brief description of two gas core reactor concepts devised for propulsion applications is presented and recent research on the emission of radiation from nonequilibrium fissioning gases is considered. The principles of UF6 and plasma core reactor experiments are discussed. It is expected that the program of nuclear gas core reactor experiments can lead in several years to reactor operation at pressures, temperatures, and power levels of technological significance to terrestrial applications.

Thom, K.↗

NASA NTRS1977

Dry coal feeder development program at Ingersoll-Rand Research, Incorporated

A dry coal screw feeder for feeding coal into coal gasification reactors operating at pressures up to 1500 psig is described. Results on the feeder under several different modes of operation are presented. In addition, three piston feeder concepts and their technical and economical merits are discussed.

Mistry, D. K.↗

NASA NTRS1983

Two-Temperature-Zone Silicon Reactor

When high purity silicon is synthesized by reduction of silicon tetrafluoride by sodium, very-fast highly exothermic reaction takes place. Controlled reaction is proposed in which SiF4-pressurized vertical reactor operates with two temperature zones. Liquid sodium feeds from nozzle at top of rector without reacting with SiF4. When sodium reaches higher temperature region at bottom, reaction takes place immediately.

Sanjurjo, A.↗