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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 formation of sparingly soluble ternary Na/Mg–Np( v )–borate(s) solid phases in alkaline, dilute-to-concentrated, NaCl and MgCl 2 solutions is confirmed by a multimethod experimental approach.
Structural, spectroscopic and theoretical analyses of mer-MBr3(OPcy3)3 (M = Am, Nd, Pr, Ce, La) reveal significant amounts of metal based p-orbital contribution.
The proton beam power limit for a solid-tungsten spallation target is largely determined by beam induced thermomechanical structural loads and decay heat power deposition, while its lifetime is limited by radiation damage and fatigue life of the target materials. In this paper, we studied the power limits of a stationary water-cooled solid tungsten target concept. Tantalum clad tungsten was considered as a reference case. Being a low activation material, zircaloy 2 cladding option was studied and its decay heat driven power limit was compared with the reference case. Zirconium alloys have proven operations records in spallation target and nuclear fission environments, supported by materials data obtained from post irradiation examinations. Recent study also demonstrated feasibility of diffusion bonding zirconium to tungsten using vanadium foil inter layer. Particle transport simulations code FLUKA was used to calculate energy deposition and decay heat power deposition in the target, based on the beam parameters technically feasible at the Second Target Station of the Spallation Neutron Source at Oak Ridge National Laboratory. The energy deposition data were used for flow, thermal, and structural analyses to determine the beam intensity limit on the target concept studied. The decay heat deposition data were used to calculate the transient temperature evolution in the tungsten volumes in a loss of coolant accident (LOCA) scenario to determine its beam power limit. For a 1.3 GeV proton beam, the power limit on a stationary target was 400 kW for a tantalum clad target model and 800 kW for a zircaloy 2 clad target model.
The conference covered the radiation effects on devices, circuits, and systems, physics and basic radiation effects in materials, dosimetry and radiation transport, spacecraft charging, and space radiation effects. Other subjects included single particle upset phenomena, systems-generated electromagnetic pulse phenomena, fabrication of hardened components, testing techniques, and hardness assurance.
A set of four tissue-equivalent proportional counters (TEPCs), with their detector heads at the centers of 0 (bare), 3, 7 and 9-inch-diameter aluminum spheres, were flown on Shuttle flight STS-89. Five such detectors at the centers of polyethylene spheres were flown 1 year earlier on STS-81. The results of dose-depth dependence for the two materials convincingly show the merits of using material rich in hydrogen to decrease the radiation exposure to the crew. A comparison of the calculated galactic cosmic radiation (GCR) absorbed dose and dose-equivalent rates using the radiation transport code HZETRN with nuclear fragmentation model NUCFRG2 and the measured GCR absorbed dose rates and dose-equivalent rates shows that they agree within root mean square (rms) error of 12.5 and 8.2%, respectively. However, there are significant depth-dependent differences in the linear energy transfer (LET) spectra. A comparison for trapped protons using the proton transport code BRYNTRN and the AP-8 MIN trapped-proton model shows a systematic bias, with the model underpredicting dose and dose-equivalent rates. These results show the need for improvements in the radiation transport and/or fragmentation models.
Fusion energy systems are currently being designed and optimized using radiation transport codes. To deal with the unique environment inside a fusion-based system, many of these designs incorporate novel materials able to withstand the high radiation fields, ensure adequate cooling and thermal protection, and produce tritium. Validation plays a vital role in building trust in the predictive power of these models and computational methods. Validation of a code consists of modeling documented real-world experiments and comparing the code-predicted response to the measured response. Adequate validation requires measured responses from real-world experiments, also known as integral data, that mimic the system being designed, including materials, impinging radiation, and temperature, among other variables. The most trusted integral data are experimental responses that have been through a rigorous benchmarking process that develops a recommended computational model and evaluates all experimental uncertainties. Finally, there are a few research groups around the world that have been producing integral data for fusion applications, but a substantial investment is needed to address the unique validation needs of the fusion community.
This report consolidates publicly available information related to pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear fuel physical characteristics and operating conditions. The intent of this report is to provide reference information for use in shielding analyses of spent nuclear fuel transportation packages and storage casks.
This report meets the requirements of M3 milestone M3SF-21OR020302066 for the Office of Integrated Waste Management (IWM) and is a revision to M3 milestone M3FT-16OR090402028 for the former Nuclear Fuels Storage and Transportation Planning Project (NFST), “Safety Record of SNF Shipments.” The US Department of Energy (DOE) has since established the IWM, which builds on the work begun by NFST, to develop an integrated waste management system for spent nuclear fuel (SNF), including the development of a large-scale transportation system for the safe transport of SNF to storage or disposal facilities.
Commercial nuclear power plants typically use nuclear fuel that is enriched to less than five weight percent in the isotope 235 U. However, recently several vendors have proposed new nuclear power plant designs that would use fuel with 235 U enrichments between five weight percent and 19.75 weight percent. Nuclear fuel with this level of 235 U enrichment is known as “high assay low-enriched uranium.” Once it has been irradiated in a nuclear reactor and becomes used (or spent) nuclear fuel, it will be stored, transported, and disposed of. However, irradiated high assay low-enriched uranium differs from typical irradiated nuclear fuel in several ways, and these differences may have economic effects on its storage, transport, and disposal, compared to typical irradiated nuclear fuel. This report describes those differences and qualitatively discusses their potential economic effects on storage, transport, and disposal.
The National Aeronautics and Space Administration’s (NASA’s) Space Nuclear Propulsion (SNP) project aims to demonstrate the feasibility of nuclear thermal propulsion (NTP) as a thrust mechanism for spacecraft. An initial Demonstration Rocket for Agile Cislunar Operations (DRACO) launch will serve as a test of the technology in Earth’s orbit. Future work may involve the development of an NTP engine and spacecraft design with the capability to transport astronauts from Earth to Mars and back safely, though NTP has other potential uses too.
Resilient operation of interdependent infrastructures against compound hazard events is essential for maintaining societal well-being. To address consequence assessment challenges in this problem space, we propose a novel policy-guided tri-level optimization model applied to a proof-of-concept case study with fuel distribution and transportation networks – encompassing one realistic network; one fictitious, yet realistic network; as well as networks drawn from three synthetic distributions. Mathematically, our approach takes the form of a defender-attacker-defender (DAD) model—a multi-agent tri-level optimization, comprised of a defender, attacker, and an operator acting in sequence. Here, in this study, our notional operator may choose proxy actions to operate an interdependent system comprised of fuel terminals and gas stations (functioning as supplies) and a transportation network with traffic flow (functioning as demand) to minimize unmet demand at gas stations. A notional attacker aims to hypothetically disrupt normal operations by reducing supply at the supply terminals, and the notional defender aims to identify best proxy defense policy options which include hardening supply terminals or allowing alternative distribution methods such as trucking reserve supplies. We solve our DAD formulation at a metropolitan scale and present practical defense policy insights against hypothetical compound hazards. We demonstrate the generalizability of our framework by presenting results for a realistic network; a fictitious, yet realistic network; as well as for three networks drawn from synthetic distributions. Additionally, we demonstrate the scalability of the framework by investigating runtime performance as a function of the network size. Steps for future research are also discussed.
The SCALE code package offers multiple nuclear data libraries supporting Monte Carlo transport, with sensitivity and uncertainty methods derived from MC transport solutions. Several libraries are multigroup, which introduce bias differing by system. Previous work has shown poor S/U results in several reflector materials: ICSBEP benchmark HMF-084 was selected to analyze biases and S/U method applicability to a variety of reflectors. Prior and ongoing work found inaccuracies in CSAS and TSUNAMI results, which were further investigated utilizing the HEU-MET-FAST-084 ICSBEP critical benchmark, chosen for its geometrical simplicity and variety of reflector materials. Perturbation of reflector thickness across various reflector materials allowed for an assortment of materials is to be tested swiftly for each sequence and method. Observation was an increasing bias of MG $k_{eff}$ relative to CE, in both direction and magnitude. IFP produced extremely reliable results. >85% of CLUTCH cases were found satisfactory.
Fusion reactors depend on the blanket material to breed and release tritium at the same rate or faster than it is consumed by the fusion reaction. Cellular ceramic breeders are dense materials that can maintain a high tritium breeding ratio while promoting tritium release thanks to highly connected pores. Assessing the tritium breeding capabilities of these materials requires a combination of extensive experimental and modeling efforts. In this work, we develop and calibrate a multiphysics model of tritium transport. This novel model accounts for ceramic and pore diffusion, trapping and detrapping, and several surface reactions at the pore surface. We perform a sensitivity analysis and calibrate the model by comparing its predictions against experimental measurements of deuterium absorption. The calibrated model is then used to model tritium absorption in samples with different pore microstructures to investigate the effect of pore interconnectivity on tritium absorption. The model is part of the development of the multiscale, multiphysics framework for tritium transport (i.e., the Tritium Migration Analysis Program [TMAP8]), which is itself built on top of the finite element multiphysics framework MOOSE. This study demonstrates some of TMAP8’s capabilities, and is the first step toward assessing the tritium breeding capabilities of ceramic breeder material designs.
Jules Horowitz Reactor (JHR) is a future center for nuclear material testing in Europe, currently under construction at CEA Cadarache in southern France. It has been modelled extensively with CEA's HORUS3D/N neutron calculation package, that includes a Monte-Carlo neutron transport code TRIPOLI-4. In order to support the future computational analyses on the reactor core physics, VTT's Monte Carlo neutronics code SERPENT 2 has been used to create an alternative model of the JHR, taking advantage of the pre-existing Constructive Solid Geometry (CSG) structures on which the TRIPOLI-4 model was built. This work describes a benchmark, in which two JHR core configurations were analyzed using both of these models. The first configuration is a fresh fuel core at the beginning of life, in which aluminum based dummy-devices are installed in the reactor test positions. The second configuration represents the end-of-cycle equilibrium core with a heterogeneous burnup distribution, in which CALIPSO devices containing cladding material samples are installed at the in-core experimental positions, while the reflector positions are occupied by DEN-REP devices containing UO{sub 2} fuel. Neutron flux spectra and heat deposition are the parameters of interest in this study, due to their high importance in creating the desired experimental irradiation environment. The impact of delayed gamma energy deposition will be studied in detail, since the two models account for it with different methods. This work aims to establish a SERPENT 2 model of JHR, understanding its differences and agreements with TRIPOLI-4, and discussing its potential future applications. (authors)
The ability to accurately model the subsurface transport of radionuclides is fundamental to the remote detection and characterization of underground nuclear explosion (UNE) events. Developing more sophisticated transport models presents a significant opportunity to enhance monitoring capabilities, particularly in the reliable prediction of signature migration. Experimentally determined characterization of geologic materials associated with transport properties is the pertinent base information for such robust model development and calibration. Here, we report results from an unprecedented study demonstrating changes to the pore and fracture network structures in geological materials in response to UNEs over nanometer to micrometer length scales. Volcanic tuffs of five different lithological formations from pre- and post-UNE environments were collected from the Nevada National Security Site. Combined ultra-small and small-angle neutron scattering techniques were used to characterize the tuff pore structure. The results demonstrate measurable differences in the specific surface area and porosity of samples pre- and post-shot from texturally similar lithological formations, indicating that pore properties can serve as a direct physical signature of a UNE. The results also provide experimentally determined transport parameters in support of advanced model development through the integration of gas migration, hydrodynamic simulations, and geologic framework models.
Radiation transport calculations are often used to estimate dose or exposure to components and personnel surrounding a radiation source. The sources for these calculations are decaying radionuclides within the various nuclear materials. Under long-term radiation exposure, some components may show signs of material damage such as hardening, cracking, discoloration, etc. resulting from the dose received from nearby radioactive sources. Radiation dose quantities are a useful means to correlate material damage or degradation to radiation interactions. The degradation is often dependent on how incident radiation interacts and deposits energy into the material. This dependence implies that radiation dose should be properly defined for the material effect of interest. Generic dose quantities, such as total energy deposited, may not necessarily correlate well for degradation of any given material.
This document is the primary deliverable for a scoping study proposed to DOE National Laboratory Announcement Number LAB 19-2114 in the NNSA research area. The study supports user applications employing active neutron interrogation by providing a science plan to improve the modeling capability and the nuclear data that radiation transport codes use. Users rely on the accuracy of the elastic scattering and non-elastic cross-sections spanning thermal energies to 14 MeV (and higher in some cases) for modeling the neutron transport through complex geometries of materials potentially composed of many elements. While the elastic scattering cross-section data are accepted for all commonly occurring elements, the non-elastic cross-section data and the associated emission data include reaction channels that require attention. The study focused on the non-elastic reactions that emit secondary, also known as prompt, gammas with the premise that many users would benefit from improved modeling of these reactions. Many users develop material assay technologies based upon gamma signatures from radiative capture, inelastic scattering, and reactions on low-Z isotopes emitting multiple particles, so the nuclear data gaps, modeling deficiencies, and recommendations for addressing the shortfalls were assessed for these reactions. Fission gammas were excluded from this study because there are other efforts underway to address known shortfalls. Follow-on efforts that successfully execute the recommendations will tangibly improve to the ability to model gamma signatures and backgrounds for user applications, such as controlled substance detection, oil-well logging, and space exploration.
This paper compares the Light-Weight Radioisotope Heater Unit (LWRHU) historical testing to Department of Transportation (DOT) Special Form criteria per 49 CFR Part 173, Shippers-General Requirements for Shipments and Packages, Section 469, Tests for special form Class 7 (radioactive) materials as specified in DOE-STD-1027, Hazard Categorization of DOE Nuclear Facilities. Compliance with the aforementioned criteria allows for the exclusion of radioactive sources from the facility’s nuclear material inventory. As part of a new program at Los Alamos National Laboratory (LANL), LWRHUs are planned to be moved into TA-55, PF-5 Power Source Assembly Area (PSAA) for assembly and testing activities. The PSAA Facility is categorized as a low hazard, non-nuclear facility per DOE STD-1027 and is described in TA55-PF-5-FHC-11-001, Facility Hazard Categorization of the PSAA at PF-5, TA-55. The Facility Hazard Categorization document states, there must be documentation that radioactive sources are engineered to pass the special form testing specified by DOT in 49 CFR 173.469, Tests for special form Class 7 (radioactive) materials or ANSI N43.6, Sealed Radioactive Sources - Classification. This paper documents the extensive historical environmental safety testing of the LWRHU and demonstrates its ability to survive potential launch safety accident scenarios. The testing results are used herein to demonstrate through a comparative analysis that the robust design of the LWRHU meets or exceeds the special form testing specifications in 49 CFR 173.469, enabling the LWRHU to be transported and handled in PF-5 in support of testing and assembly program.