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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Wettability and CHF limits of Accident-Tolerant nuclear fuel cladding materials in light water reactor conditions

Here, we present the results of experimental investigations aimed at evaluating the thermal–hydraulic performance of chromium-coated zircaloy, i.e., one of the most promising accident tolerant fuel (ATF) cladding material for light water nuclear reactors. Precisely, we investigate the wettability and critical heat flux (CHF) limits of chromium-coated and conventional zircaloy surfaces in prototypical reactor conditions. For both surface types, we measure the contact angle in a vapor-saturated atmosphere from ambient pressure to the operating pressure of pressurized water reactors (PWRs), i.e., ~15 MPa. We measure the ambient-pressure steady-state flow boiling CHF with a spatially uniform heat flux. We measure the high-pressure steady-state flow boiling CHF with a cosine shape heat flux (up to 20 MPa) and with a uniform heat flux (up to 15 MPa), also exploring the effect of CRUD deposits on the chromium-coated surface. Our results reveal that the chromium surface and the bare zircaloy surface have similar wettability and both become super-hydrophilic in PWR conditions, and that there is practically no difference in the steady-state CHF limits, both at low-pressure and high-pressure conditions, also when the chromium-coated surface is covered by a CRUD deposit. However, while the chromium-coating does not improve the CHF compared to the bare zircaloy surface, it improves the post-CHF behavior. The chromium coating prevents the reaction between zircaloy and steam, which results in the formation of a brittle zirconium oxide through the surface of the cladding. We also measure the transient CHF under exponentially escalating heat flux inputs of a nano-smooth and a rough surface mimicking a chromium-coated zircaloy cladding. Interestingly, the results of the transient heat flux tests suggest that the CHF limit for very short periods (i.e., fast transients) is independent of the surface finish, being the same for a rough chromium surface or a nano-smooth surface.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Water vapor oxidation behaviors of nuclear graphite IG-110 for a postulated accident scenario in high temperature gas-cooled reactors

Water leakage in accidental conditions of high temperature gas-cooled reactors is one of the most critical problems that can compromise the integrity of different nuclear components. In this study, oxidation behaviors of nuclear graphite IG-110 in water ingress accidental conditions were investigated. Mass loss and oxidation rates were evaluated after oxidation tests at temperatures up to 1400 °C in an Ar-20 vol% H 2 O mixed atmosphere. The activation energy decreased from 318.6 to 148.9 kJ/mol with temperature, indicating two different oxidation regimes. The cross-sections of the oxidized samples were systematically characterized. The corresponding logarithmic porosity profiles showed a temperature dependency. Pore formation moved toward near-surface regions with increasing temperature and preferential binder oxidation, with filler particle degradation. Furthermore, oxidant concentration profiles and oxidation depths were estimated using a theoretical model and compared with the experimental results. This work provides important benchmark data and safety analysis guidance for the accident scenario in high temperature gas-cooled reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

U 3 Si 2 and UO 2 composites densified by spark plasma sintering for accident-tolerant fuels

This work reports the synthesis and characterization of the U 3 Si 2 and UO 2 composites sintered by spark plasma sintering (SPS) with controlled microstructures for accident-tolerant fuel application. The U 3 Si 2 and UO 2 composites with different silicide and oxide ratios were synthesized by SPS at temperatures from 1000 to 1300 °C for 5 minutes. The microstructure and phase composition of the SPS densified composite fuels were characterized with scanning electron microscopy, X-ray diffraction (XRD), and energy dispersed spectroscopy (EDS). A systematic study of the thermal and mechanical properties was conducted using microhardness testing and laser flash apparatus, along with oxidation resistance measurements using thermogravimetric analysis (TGA). The results show that the synthesis of composite fuels can be achieved with a 90% theoretical density (TD) at 1000 °C and over 95% TD when sintered at 1300 °C. XRD and EDS results confirmed that the dominant phases in the composites are U 3 Si 2 and UO 2 . Improved physical density generally leads to improved hardness, fracture toughness, thermal diffusivity, and onset temperature during the oxidation process. U 3 Si 2 was found to play a dominant role in determining the mechanical and oxidation properties of the composite fuels, whereas UO 2 had a more important impact on controlling the thermal diffusivity of the composites. The composite with 50 wt% UO 2 sintered at 1300 °C displayed the onset oxidation temperature of 500 °C by dynamic oxidation testing using TGA at a ramp degree of 10 °C/min. The composite also achieved a high fracture toughness of ~3.5 MPa m ½ . Finally, these results highlight the potential of composite fuel forms densified by SPS with simultaneously enhanced fissile element density, fracture toughness, thermal transport properties, and oxidation resistance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

X-ray and gamma-ray tomographic imaging of fuel relocation inside sodium fast reactor test assemblies during severe accidents

The present work reports on x-ray and gamma-ray high-spatial resolution computerized tomography measurements of the Pin Bundle Metallic Fuel Relocation (PBR) assemblies tested in the Metallic Uranium Safety Experiment (MUSE) facility at Argonne National Lab (ANL). The aim of the study was to characterize fuel relocation structures that develop during severe core accidents pertaining to SFR assemblies; these include but are not limited to advanced core disruption recreated in the PBR-1 assembly, and cladding breach recreated in the PBR-2 assembly. We report the x-ray tomography measurements were able to resolve small quantities of relocation fuel; with increased presence of relocation fuel, the x-ray 1measurements spatially mapped the material but could not resolve the inner regions of these. The gamma-tomography measurements showed improved results, resolving the relocation structures in great detail. The upper plenum of the PBR-1 assembly where the molten uranium was initially inserted presented high structural damage, reflected by the partial and complete disintegration of the central rods. Relocation fuel filled the subchannels, adhering to surviving cladding walls and the assembly casing. In the lower portion of the measured section, the tomogram degrades due to photon starvation effects hinting at the increased amount of relocation fuel potentially plugging the assembly; flow blockage in this section was difficult to determine due to the tomogram’s degradation from photon starvation. Small fragments were observed further down the assembly, dislodged from the initial insertion of the molten material. This section was used as an unperturbed assembly reference, with a calculated blockage of less than 1% from the present fragments. The PBR-2 assembly was characterized by columnar relocation structures propagating through the subchannels. Three of the relocation structures were captured in the measured section, with evidence of cross migration on to adjacent subchannels. The measured section in this assembly captures the leading edge of two structures. The calculated flow blockage was 16% in the planes where the three relocation structures are present, but this quickly decreases to approximately 5% past the leading edge of two of the structures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Irradiation-induced amorphization of Fe-Y-based second phase particles in accident-tolerant FeCrAl alloys

Here, second phase intermetallic particles in an advanced accident-tolerant FeCrAl (Fe-13Cr-5Al-2Mo) alloy are formed in the α-Fe matrix during processing. These particles are prominently related to the added Y. Neutron irradiation to ~7 displacements per atom (dpa) with a dose rate of ~8.16 × 10 -7 dpa/s at 282 °C resulted in the amorphization of these precipitates which could degrade the mechanical properties of the FeCrAl alloys. Analytical electron microscopy and diffraction analysis combined with structural freedom analysis have been used to investigate the radiation resistance of the second phase particles. Radiation tolerance is closely linked to the particle Fe-Y content and can be tailored using the structure freedom value.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Sensitivity analysis, surrogate modeling, and optimization of pebble-bed reactors considering normal and accident conditions

This research provides a valuable tool that streamlines the optimization process while significantly increasing its accuracy. This study creates a robust framework for reactor design optimization by incorporating comprehensive modeling using the Comprehensive Reactor Analysis Bundle, or BlueCRAB, within the Multiphysics Object-Oriented Simulation Environment (MOOSE). BlueCRAB is the United States Nuclear Regulatory Commission's code suite for non-light water reactor analysis and includes the Griffin, Pronghorn, and Bison applications. This not only improves the efficiency of the optimization process but also enhances the reliability of the results. Such a tool is essential for advancing the state-of-the-art in pebble-bed reactor technology and is critical for achieving the goals of Generation IV reactors, which aim for safe, sustainable, and economically viable nuclear energy solutions. This work presents and applies this workflow on pebble-bed reactors while considering both normal and off-normal conditions. A representative gas-cooled pebble-bed reactor at equilibrium core conditions serves as the nominal design specification for normal operation and is based on previous research. The depressurized loss-of-forced-cooling accident is deployed for off-normal conditions in this work. After defining design-related parameters and quantities of interest regarding reactor safety and performance, this multiphysics model is sampled using the MOOSE stochastic tools module. The result is a comprehensive dataset of configurations, enabling sensitivity analysis and the generation of surrogate models. Subsequently, the dataset and surrogate models are employed in two optimization studies aimed at maximizing fuel utilization and economic profit while adhering to safety and operational constraints. Performing the optimization process with fuel utilization as the metric leads to an improvement of approximately 10%, compared to engineering-judgment-based nominal conditions. The optimization on economic profit leads to an estimated increase of ~300 million USD over the lifetime of the reactor.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Chemical characterisation of degraded nuclear fuel analogues simulating the Fukushima Daiichi nuclear accident

The Fukushima Daiichi accident generated degraded nuclear fuel material, mixed with other reactor components, known as molten core-concrete interaction (MCCI) material. Simulant MCCI material was synthesised, excluding highly radioactive fission products, containing depleted U, and incorporating Ce as a surrogate for Pu. Multi-modal µ-focus X-ray analysis revealed the presence of the expected suite of U-Zr-O containing minerals, in addition to crystalline silicate phases CaSiO 3 , SiO 2 -cristobalite and Ce-bearing percleveite, (Ce,Nd) 2 Si 2 O 7 . The formation of perclevite resulted from reaction between the U-Zr-O-depleted Ce-Nd-O melt and the silicate (SiO 2 ) melt. It was determined that the majority of U was present as U 4 , whereas Ce was observed to be present as Ce 3+ , consistent with the highly reducing synthesis conditions. A range of Fe-containing phases characterised by different average oxidation states were identified, and it is hypothesised that their formation induced heterogeneity in the local oxygen potential, influencing the oxidation state of Ce.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Analysis of Postulated Accident Scenarios for the Transformational Challenge Reactor

The Transformational Challenge Reactor (TCR) is a helium-cooled, yttrium-hydride-moderated reactor that was designed for the U.S. Department of Energy Office of Nuclear Energy. A key objective of the TCR was to employ advanced manufacturing techniques in a nuclear system and demonstrate their potential for revolutionizing the nuclear reactor design process. One purpose of the present work is to demonstrate the safety of the TCR under postulated accidents. Based on RELAP5-3D and COMSOL analyses, the TCR remained below all current safety limits and far below the expected failure limits for the core materials. Another purpose of this work is to provide useful insights and recommendations regarding the application of RELAP5-3D to gas-cooled or other advanced reactors. A novel approach was implemented for simultaneously modeling conduction and radiation in RELAP5-3D, which was found to provide reasonable predictions of radial core, vessel, and ex-vessel heat transfer during postulated events. A multicode approach was also applied, in which high-fidelity COMSOL calculations were used to tune the radial heat transfer parameters in RELAP5-3D. The tuned RELAP5-3D model demonstrated comparable peak temperature predictions as COMSOL, despite a coarse treatment of the core in RELAP5-3D consisting of only two lumped heat structures. This high-fidelity tuning approach enabled enhanced accuracy as well as minimal complexity within the RELAP5-3D model, even for complex fuel geometric designs as in the TCR. Finally, investigations were made into the potential for flow reversal during a pressurized loss-of-forced-flow event in the TCR. The TCR is designed with downward helium flow through the core during normal operation. The RELAP5-3D model predicted that this downward flow would persist, without flow reversal, up to several days after the circulator trip. This was attributed to natural circulation hysteresis effects as have been noted in similar thermofluidic systems. Although flow stagnation and eventual reversal did not lead to unsafe TCR conditions, interesting spatial effects were observed which may have safety relevance for other reactor system designs and coolant types that are designed for downward core flow during normal operation, warranting closer investigation of the flow reversal phenomenon.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Application of MELCOR for Simulating Molten Salt Reactor Accident Source Terms

Molten Salt Reactor (MSR) systems can be divided into two basic categories: liquid-fueled MSRs in which the fuel is dissolved in the salt, and solid-fueled systems such as the Fluoride-salt-cooled High-temperature Reactor (FHR). The molten salt provides an impediment to fission product release as actinides and many fission products are soluble in molten salt. Nonetheless, under accident conditions, some radionuclides may escape the salt by vaporization and aerosol formation, which may lead to release into the environment. Here, we present recent enhancements to MELCOR to represent the transport of radionuclides in the salt and releases from the salt. Some soluble but volatile radionuclides may vaporize and subsequently condense to aerosol. Insoluble fission products can deposit on structures. Thermochimica, an open-source Gibbs Energy Minimization (GEM) code, has been integrated into MELCOR. With the appropriate thermochemical database, Thermochimica provides the solubility and vapor pressure of species as a function of temperature, pressure, and composition, which are needed to characterize the vaporization rate and the state of the salt with fission products. Since thermochemical databases are still under active development for molten salt systems, thermodynamic data for fission product solubility and vapor pressure may be user specified. This enables preliminary assessments of fission product transport in molten salt systems. In this paper, we discuss modeling of soluble and insoluble fission product releases in a MSR with Thermochimica incorporated into MELCOR. Separate-effects experiments performed as part of the Molten Salt Reactor Experiment in which radioactive aerosol was released are discussed as needed for determining the source term.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

SCALE Shielding Calculations for Advanced Reactor Accident Scenarios

The study results presented in this paper demonstrate the capabilities of the SCALE computer code for non–light water reactor (non-LWR) radiation source term and shielding calculations in support of the U.S. Nuclear Regulatory Commission non-LWR fuel cycle demonstration project. Representative non-LWR types, including the sodium-cooled fast reactor (SFR), the molten salt reactor (MSR), and the heat pipe microreactor (HPMR), were analyzed to evaluate dose rates associated with postulated accident scenarios (SFR and MSR), reactor operation (MSR and HPMR), and fuel self-protecting characteristics (HPMR). New features were implemented in SCALE depletion codes to better simulate MSR operation.

MSR↗
DOE OSTI2026

Multiphysics Analysis of Li Cooled Divertor Substrate During Loss of Coolant Accident (LOCA)

In the ongoing study of potential designs for liquid metal (LM) plasma-facing components (PFCs), so-called “slow” and “fast” Li flow divertor concepts are under investigation. In the previous studies on design and analysis of the slow Li flow divertor and comparison with the fast Li flow divertor, the magnetohydrodynamics (MHD)/heat transfer effects of the Li flowing inside the substrate as a second coolant were comprehensively investigated under the normal steady-state operation conditions. Here, in the present study, the multiphysics analysis is extended to the unsteady abnormal divertor scenario where the Li layer on top of the substrate does not provide full coverage or even totally disappears for a certain period of time, so that the substrate becomes directly exposed to the incident high plasma heat flux. Such an unwanted event may happen regardless of the concept of the divertor and is worth detailed investigations, typically referred to as a loss of coolant accident (LOCA). To address this situation, a simplified scoping analysis is conducted first in 2-D, and then an integrated 3-D modeling is performed using a time-dependent multiphysics model in COMSOL Multiphysics that integrates LM MHD, heat transfer, and solid mechanics. The main goal is to evaluate conditions under which the major material limits, such as the maximum allowable temperature, stress, and displacement of the substrate, can still be met. It was shown that the maximum time over which the substrate of RAFM steel can retain structural integrity during the LOCA is around 0.2 ∼ 0.3 s. Any divertor concept that utilizes RAFM steel as a substrate material and liquid Li as a second coolant should take such a permitted time into consideration.

divertor↗
DOE OSTI2020

Criticality Accident Response

The document outlines procedures and priorities for Radiation Protection personnel response should a nuclear criticality accident occur.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

Scan of an Unpublished Report: "Fission Product Behavior During Severe LWR Accidents: Recommendations for the MELCOR Code System"

This document provides a scanned version of a 1987 SAND report that was never formally published. However, this report was referenced within the MELCOR Reference Manual and, therefore, provides historical information and technical basis for the MELCOR code. This document is being made available to permit users of the MELCOR code access to the information. The title page has been edited to prevent any confusion with regards to the possible documentation identifiers, such as the SAND report number or the intended date of publication. Beyond these modifications, a cover, distribution, and back cover are prepended and appended to the document to conform to modern SAND report style guidelines. The first four chapters of this report were updated and released under the title "Fission Product Behavior During Severe LWR Accidents: Recommendations for the MELCOR Code System. Volume I" and were made available by the U.S. NRC through the Adams database under accession number ML19227A327. No prior release of the remaining content of this report has occurred.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

A Coupled Experimental and Simulation Approach to Investigate the Impact of Grain Growth, Amorphization, and Grain Subdivision in Accident Tolerant U 3 Si 2 Light Water Reactor Fuel

Triuranium disilicide (U 3 Si 2 ) is being considered as an alternative accident tolerant fuel (ATF) due to its higher thermal conductivity. However, there is uncertainty in its use due to a lack of irradiation data at light water reactor (LWR) conditions. In this project, we used an integrated experimental and simulation approach to answer two significant questions pertaining to U 3 Si 2 that were not investigated by the Fuel Cycle Research and Development (FCRD) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) programs: (1) Will grain growth in the hotter portions of the fuel significantly impact U 3 Si 2 LWR fuel behavior? (2) Under what conditions do grain subdivision and amorphization occur in U 3 Si 2 and will either occur at LWR conditions? Our approach to answer these questions used in situ ion irradiation and annealing to investigate grain growth, grain subdivision, and amorphization, along with mesoscale simulations using the MARMOT tool. We found that while the grain boundary mobility of U 3 Si 2 is larger than that of UO 2 , grain growth in U 3 Si 2 fuel pellets will be less than in UO 2 and may be neglected due to the lower thermal conductivity and thus lower centerline temperature. We also found that amorphization will not occur in U 3 Si 2 above 600 K, such that it is not likely to occur at typical LWR operating conditions. Grain subdivision will occur at LWR conditions and will likely occur at lower burnups than in UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

High Temperature Steam Oxidation of Irradiated FeCrAl in the Severe Accident Test Station

FeCrAl-UO2 test capsules were fabricated at Oak Ridge National Laboratory (ORNL) and irradiated at the Advanced Test Reactor. Following irradiation, samples were sectioned from the irradiated rod and oxidation kinetics were evaluated to access the candidate cladding high-temperature oxidation performance following irradiation. The irradiation was performed at approximately 400°C to a burnup of 10 GWd/MT. The high-temperature oxidation tests were conducted in the ORNL Severe Accident Test Station at 1200 and 1300°C. Weight measurements were taken before and after oxidation testing. Cross-sections of the cladding were metallographically mounted and optical microscopy was performed. Measurements of the oxidation layer before and after high-temperature testing were collected. The results indicate the irradiated FeCrAl C35M alloy provided good thermal stability up to 1200°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

High Level Gap Analysis for Accident Tolerant and Advanced Fuels for Storage and Transportation

This initial gap analysis considers proposed accident tolerant fuel (ATF) options currently being irradiated in commercial reactors, since these are most likely for future batch implementation. Also, advanced fuel (AF) options that may be likely for use in advanced reactors are considered. The cladding technologies considered were chromium-coated zirconium-based alloys, FeCrAl, and both monolithic and matrix composite Silicide carbide (SiC). The fuel technologies considered were chromium-doped uranium dioxide fuel, uranium alloys, uranium nitride, and uranium silicide. Numerous national labs, industry, and countries are performing significant testing and modeling on these proposed technologies to establish performance, but at this time none of the prototypes being irradiated have achieved end-of-life (EOL) burnup. There are some testing results after one burnup cycle to verify in-reactor performance, but little data beyond that. As the ATF prototypes acquire more burnup, data will be produced that is relevant to storage and transportation. The DOE:NE Spent Fuel and Waste Science and Technology (SWFST) Storage and Transportation (ST) Control Account will evaluate the performance data as it becomes available for application to the identified gaps for ST.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

SCALE Analysis of a Fluoride Salt-Cooled High-Temperature Reactor in Support of Severe Accident Analysis

As part of a US Nuclear Regulatory Commission–sponsored project to assess the modeling and simulation capabilities for accident progression, source term, and consequence analysis for advanced reactor technologies with SCALE and MELCOR, SCALE was used for the modeling and simulation of a fluoride salt-cooled high-temperature reactor (FHR). Based on the preconceptual design for a small modular 236 MWth FHR developed by the University of California, Berkeley (PB-FHR-Mk1), a SCALE model of the PB-FHR-Mk1 reactor core was developed. The reactor was modeled at equilibrium state with different fuel compositions in different regions of the reactor. An iterative approach was used to interpolate and mix the burnup-dependent fuel compositions obtained through the depletion calculation of a core slice model. After demonstrating the applicability of SCALE’s multigroup (MG) approach for the simulation of the PB-FHR-Mk1, the resulting equilibrium core was studied in terms of the power profile, the flux profile, temperature reactivity coefficients, and the xenon reactivity. Furthermore, the tritium production rate in the salt coolant was determined, and the dependence of the one-group cross sections on the burnup and location in the reactor core was studied. The results obtained with SCALE were post-processed to provide the MELCOR team with the core inventory and decay heat of the equilibrium core, a zone-wise power profile, temperature feedback coefficients, the tritium production rate, and the xenon worth.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Universal Nuclear Accident Dosimeter (UNAD) User Group Meeting - June 1, 2022

A brief summary was provided of the proposal in Reference (a), which motivated the formation of this user group. The project is funded for four years (FY 2022-2025); the User Group will be a resource through the four years. The group will meet periodically to share knowledge, experience, and expertise to work towards the development of an improved and universal nuclear accident dosimeter. Members were welcomed and each introduced themselves.

61 RADIATION PROTECTION AND DOSIMETRY↗