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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Kinetics and transport of hydrogen in graphite at high temperature and the effects of oxidation, irradiation and isotopics

The kinetics of uptake and desorption impact the performance of graphite as a vector for tritium in high-temperature fission reactors and in the blanket of fusion reactors. Graphite components in these reactors are exposed to temperatures > 500 °C and H 2 partial pressures of few Pa and desorption temperatures are limited to < 1600 °C; limited data is available at these conditions. Here we review the mechanisms for uptake in, transport and desorption of hydrogen from graphite at high temperature, compiling data on uptake rates, diffusion coefficients and activation energies and providing a discussion of the impact of irradiation, pre-oxidation and isotope. At FHR conditions, trapping impacts uptake rates, leading to a reduction in apparent diffusivity by 35 to 80% compared to higher partial-pressure uptake. Timelines for desorption are not clearly defined; extrapolating from available data, at 1150 °C desorbing 80% of tritium uptaken at FHR conditions may take from 100 to 10,000 h.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Tritium generation, release, and retention from in-core fluoride salt irradiations

Further understanding of tritium transport mechanisms in the combined molten fluoride salt and graphite environment is necessary for the design and licensing of a Fluoride-Salt-Cooled High-Temperature Reactor (FHR). The three in-core fluoride salt irradiations completed at the Massachusetts Institute of Technology Reactor (MITR) are a useful parallel for studying transport phenomena expected in a FHR environment. During the irradiations, evolution of tritium from the flibe salt was monitored and compared to the calculated total generation rate. A difference of 22 ± 10% between the integrated calculated tritium generation rate and the total release was measured for the third MITR irradiation (FS-3). The fraction of tritium which was not released from the salt could be explained by tritium retention in graphite. Additionally, for post irradiation examination, a thermal desorption furnace was used to heat nuclear graphite samples in order to release and measure retained tritium. The desorption analysis in this work utilized seven subsections of graphite from the second salt irradiation (FS-2); three from a disc of IG-110U and four from ARB matrix graphite. Observed desorption versus temperature as well as total tritium content in the samples after irradiation indicate that the graphites were not volumetrically saturated with tritium, but rather tritium retention was likely limited to the near-surface region. Measurements of the samples resulted in 2.90 ± 0.29 μCi/mm 2 of tritium retained by IG-110U and 1.83 ± 0.31 μCi/mm 2 for ARB during the 300 h FS-2 in-core irradiation. Based on the desorption measurements, the estimated total tritium retention in graphite from the FS-2 samples is consistent with the tritium release measurements from the FS-3 experiment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Verification and Demonstration of One-Dimensional Freezing Model in SAM for Salt-Cooled Reactor Analysis Applications

This work presented the development and implementation of the one-dimensional freezing model in system analysis code, SAM, as well as code verification, and code demonstration during a postulated overcooling transient, for fluoride salt-cooled high-temperature reactor (FHR) system and safety analysis applications. The paper at first summarized the freezing model, finite element numerical method, and special numerical treatment for handling phase appearance/disappearance. Analytical solutions were derived for two cases (with and without solid walls) for code verifications purpose. As expected, numerical results predicted by the SAM code agreed very well with the analytical solution. A code demonstration was then performed on a postulated protected overcooling event transient of a generic reference PB-FHR design. The code was found to successfully predict salt freezing during such a postulated event. However, due to lack of salt freezing testing data, code validation has not been performed in this work, which will be pursued in later studies when such data becomes available.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

SCALE Input and Result Files Supporting SCALE Inventory and Reactivity Analysis as Part of the Hermes 2021 PSAR Review

This dataset contains input and result files of computational simulations with the SCALE code system. The simulations cover radionuclide inventory and reactivity analyses of a fluoride salt-cooled high temperature pebble-bed reactor (PB-FHR), specifically the Hermes low-power PB-FHR demonstration reactor. Users wanting to reproduce results from this dataset are required to obtain a license to the SCALE code system for which details on the distribution can be found here: https://www.ornl.gov/scale/releases

equilibrium core↗
DOE OSTI2025

Graphite waste classification and disposal cost estimation for high temperature gas and salt reactors

As high-temperature reactor designs progress to demonstration, managing the radioactive wastes from these systems presents unique challenges. This work explores the irradiated graphite source term produced by three reactor designs: The Modular High Temperature Gas reactor (MHTGR), a pebble-bed High Temperature Gas Reactor (pb-HTGR), and a Fluoride-cooled High-temperature Reactor (FHR). We predicted a C-14 concentration of 4.3 Ci/m 3 for the MHTGR, 1.2 Ci/m 3 for the pebble bed HTGR, and 2.5 Ci/m 3 for the gFHR after 20 years of operation. The final C-14 concentration highly depended on the graphite nitrogen impurity, a major precursor for C-14. The C-14 concentration in all reactor types exceeded the 0.8 Ci/m3 threshold, resulting in a Class C waste classification. The costs associated with accepting the graphite after 20 years in a low-level waste disposal facility were projected to be 255 dollars per kWe for the MHTGR, 248 dollars per kWe for the pb-HTGR, and 56.8 dollars per kWe for the FHR.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Heat pump water heater enhanced with phase change materials thermal energy storage: Modeling study

A promising solution to improve the first hour rating (FHR) of a heat pump water heater (HPWH) involves employing a secondary tank which contains phase change material (PCM) capsules. To better understand the influence of PCM thermal storage on the HPWH operational performance, a dynamic model was developed to simulate and analyze the behavior of a newly developed HPWH technology that incorporates PCM storage into a standard HPWH to optimize key parameters such as the uniform energy factor and FHR. Mathematical models of several key components of the proposed HPWH-PCM integrated thermal energy storage (TES) system, e.g., water heater tank, PCM TES tank, evaporator, compressor, and expansion valve, have been elaborated. Also, a model-based control co-simulation platform was developed to integrate a embed PCM storage HPWH dynamic model with a control model for better supporting control design, analysis, verification, and validation. The model accuracy has been validated through comparing simulation results with lab test results, with a mean average percentage error of <5.5% for most of the selected performance variables. In addition, using the developed co-simulation platform, the demand response control strategy was studied to evaluate the load flexibility of the combined HPWH-PCM storage system by shifting system power usage to outside of the 3.5 h peak load period.

25 ENERGY STORAGE↗
DOE OSTI2021

Parametric Studies of Heat Pump Water Heater Using Low GWP Refrigerants

A storage heat pump water heater (HPWH) comprises a vapor compression system and a hot water tank. It can lead to more than 70% energy saving, compared with electric resistance heating. HPWH usually uses a wrapped-tank coil as the water heater. The wrapped coil has a strong interaction with the water tank and impose a new challenge for the engineering design and simulation study. Based on the US DOE/ORNL Heat Pump Design Model (HPDM), i.e., a public-domain, component-based thermal system modelling platform, a segment-to-segment wrapped-tank coil model coupled with a one-dimensional stratified water tank model were developed, which facilitates modeling and design HPWHs having various configurations and components. Additionally, the hardware-based design model was used to evaluate refrigerants having lower global warming potentials to replace R-134a, including R-1234yf, R-1234ze, R-290, R-513A and R-450A, when effects of compressor size, refrigerant charge on 24-hour uniform energy factor (UEF) and first-hour rating (FHR) were investigated.In the drop-in replacements, R-1234ze achieved the best UEF, and R-290 resulted in the lowest UEF. When downsizing the compressor to maximize the UEFs, the UEFs of R-134a, R-290, R1234yf, R-450A and R-513A ascended to the range of 4.4 to 4.53, with R-513A having the highest UEF. The R-1234ze ended up having the lowest UEF of 4.27. Refrigerants having larger volumetric vaporization heat led to higher FHRs if using the fixed-size compressor, that R-290 resulted in the highest FHR.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2023

Simulation of a TRISO MiniFuel irradiation experiment with data-informed uncertainty quantification

An irradiation experiment using tristructural isotropic (TRISO) fuel particles and the miniature fuel (MiniFuel) irradiation vehicle was performed in Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR) to support development of the Kairos Power fluoride salt–cooled, high-temperature reactor (KP-FHR). Here, this paper describes modeling predictions of temperatures and fuel burnup for the as-built experiment. An uncertainty quantification (UQ) analysis was performed to determine the effect of TRISO particle volume and position on the temperature predictions at various fuel heat generation rates (HGRs). This UQ study utilized fuel kernel position and volume measurements previously collected using X-ray computed tomography (XCT) techniques and Monte Carlo sampling methods to generate fuel compact cases that were then analyzed using a finite element thermal model. The UQ analysis indicated that uncertainty in calculated temperatures caused by varying TRISO particle arrangement is relatively small, even at high fuel HGR. Final predictions of particle temperatures throughout the irradiation are shown to be relevant to KP-FHR normal and off-normal operating conditions and to previous TRISO irradiation experiments. The combination of XCT with UQ analyses will inform post-irradiation examination (PIE) of the irradiated fuel compacts, and these analyses can be used to develop fuel performance models for coated particle fuel forms. Both PIE of separate-effects irradiation data and enhanced fuel performance modeling support accelerated qualification of TRISO fuels for a broad range of advanced reactor applications. The novel approach demonstrated here of measuring TRISO particle configurations with XCT methods and generating representative fuel compacts for finite element modeling and UQ analysis could be leveraged by the broader particle fuel community in the development of other TRISO fuel experiments in which these variables may have a significant impact on key outcomes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Development and Experimental Validation of a Heat Transfer Model for Spilled Molten Salt Pools

A spill of radionuclide-bearing molten salt is one of the major postulated events that needs to be analyzed for liquid fluorine salt-cooled high-temperature reactor (FHR) or molten salt reactor licensing purposes. In this postulated event, radioactive source term materials (RSTMs) in the molten salt are discharged from the reactor vessel to the reactor building. The release of RSTMs from the spilled salt pool to the gas space in the reactor building is expected to be controlled by the cooling behavior of the spilled salt, including the growth and shrinkage of the solid crust on the surface of the spilled salt pool. This paper presents a simulation model for spilled salt pool heat transfer and validation efforts. The validation data come from two molten salt spill tests that were performed recently: the PELE2 test by the Rapid Experimental Laboratory of Kairos Power LLC (KP) and the Argonne salt cooling test conducted by Argonne National Laboratory. The former was a large-scale test involving kilograms of molten spilled FLiNaK salt, and the latter was a relatively smaller-scale test targeting various processes associated with a salt spill event. Both tests generated valuable data sets that can be used to assess salt cooling and validate evaluation models. This paper provides a new one-dimensional model that can simulate the cooling process of a spilled salt pool as well as the thermal responses of heat structures, such as the stainless steel liner and the concrete below the salt. The model has been implemented as part of KP-SAM code, which is a branch of the systems code SAM specific to KP FHR. In conclusion, the simulation results of the model are compared with the data of the PELE2 and Argonne tests, and reasonable agreements are observed between the model and test data.

heat transfer model↗
DOE OSTI2025

SCALE Input and Result Files Supporting SCALE Inventory and Reactivity Analysis of the gFHR

This dataset contains input and result files of computational simulations with the SCALE code system. The simulations cover radionuclide inventory and reactivity analyses of a fluoride salt-cooled high temperature pebble-bed reactor (PB-FHR), specifically the generic FHR benchmark. Users wanting to reproduce results from this dataset are required to obtain a license to the SCALE code system for which details on the distribution can be found here: https://www.ornl.gov/scale/releases

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Neutronics and thermal-hydraulics simulation of generic pebble-bed fluoride-salt-cooled high-temperature reactor (gFHR)

The fluoride-salt-cooled high-temperature reactor (FHR) is one type of the advanced reactors and has been attracting great interest from the research institutes and commercial companies in the recent years. However, currently the technology is relatively immature. To facilitate the design and safety analysis of FHRs, Kairos Power published the generic FHR (gFHR) benchmark. In this paper, a hybrid method combining the stochastic code and deterministic code is developed to simulate the gFHR benchmark. Serpent 2 is employed as a few-group cross section generator and the cross sections are applied to the finite difference neutron diffusion code AGREE. A good consistency with the gFHR benchmark is achieved. The agreement between the Serpent 2 results and the AGREE results shows that the hybrid method is applicable to FHRs. The thermal-hydraulics are coupled to neutronics for the steady-state calculations and good agreement between AGREE and SAM is achieved. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Thermochemical Modeling in Molten Fluoride Salts for Radionuclide Speciation

An important aspect of the licensing process for nuclear reactors is providing a reasonable assurance of safety to the general public. This includes modeling potential radionuclide releases from the reactor during normal operations and accident scenarios, which is known as the reactor’s source term. A new class of advanced (non-LWR) reactors are being developed which use molten salts as the coolant fluid. Because the molten salt coolant represents a credited barrier for radionuclide transport between the fuel and the environment, a necessary aspect of mechanistic source term (MST) analysis for the KP-FHR is modeling the thermochemistry of molten salts. Provided here is a review of the theory of the thermodynamic principles governing multicomponent phase equilibria, the background of molten salt thermochemistry research, and a summary of the thermochemical data relevant to the KP-FHR coolant salt, Li 2 BeF 4 , commonly referred to as “FLiBe”. A review of literature is followed by a brief introduction to methods that can be used to model the thermochemical behavior of molten salt mixtures. The methodology outlined is based on the use of a commercial thermodynamic modeling software called FactSage, which is one of only a few available softwares based on the modified quasichemical model(MQM), which is the recommended solution model for molten salts.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Post-Irradiation Examination on MiniFuel UCO and UO 2 TRISO Particles Irradiated in HFIR at High Power

Post-irradiation examination (PIE) of MiniFuel compacts was conducted at Oak Ridge National Laboratory (ORNL) under the Nuclear Science User Facilities project in collaboration with Kairos Power (KP) to evaluate the performance of tristructural-isotropic (TRISO) particles under high particle power and fluoride-salt-cooled high-temperature reactor (FHR)-relevant conditions. MiniFuel compacts containing low-enriched uranium oxide-uranium carbide (LEUCO), low-enriched uranium dioxide (LEUO2), and natural UCO (NUCO) kernels were irradiated for four cycles at ORNL’s High Flux Isotope Reactor (HFIR) at target temperatures between 500°C and 900°C. Post irradiation, the experiment was disassembled at ORNL to recover the MiniFuel subcapsules, which were subsequently punctured to measure fission gas release. Subcapsule disassembly allowed the recovery of components of interest, such as silicon carbide (SiC) thermometry, fuel specimens, fission product sinks, and SiC spacers. The experimental irradiation temperature was confirmed by analyzing the SiC thermometry via dilatometry. PIE on the fuel specimens included gamma spectrometry and deconsolidation leach burn leach, which were complemented by imaging techniques such as x-ray computed tomography, optical microscopy, and electron microscopy. The PIE results provide insight into TRISO particle integrity, fission product retention, coating performance, and kernel migration, informing fuel qualification for application in KP’s FHR concept.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Post-Irradiation Examination on MiniFuel UCO and UO 2 TRISO Particles Irradiated in HFIR at High Power

Post-irradiation examination (PIE) of MiniFuel compacts was conducted at Oak Ridge National Laboratory (ORNL) under the Nuclear Science User Facilities project in collaboration with Kairos Power (KP) to evaluate the performance of tristructural-isotropic (TRISO) particles under high particle power and fluoride-salt-cooled high-temperature reactor (FHR)-relevant conditions. MiniFuel compacts containing low-enriched uranium oxide-uranium carbide (LEUCO), low-enriched uranium dioxide (LEUO 2 ), and natural UCO (NUCO) kernels were irradiated for four cycles at ORNL’s High Flux Isotope Reactor (HFIR) at target temperatures between 500°C and 900°C. Post irradiation, the experiment was disassembled at ORNL to recover the MiniFuel subcapsules, which were subsequently punctured to measure fission gas release. Subcapsule disassembly allowed the recovery of components of interest, such as silicon carbide (SiC) thermometry, fuel specimens, fission product sinks, and SiC spacers. The experimental irradiation temperature was confirmed by analyzing the SiC thermometry via dilatometry. PIE on the fuel specimens included gamma spectrometry and deconsolidation leach burn leach, which were complemented by imaging techniques such as x-ray computed tomography, optical microscopy, and electron microscopy. The PIE results provide insight into TRISO particle integrity, fission product retention, coating performance, and kernel migration, informing fuel qualification for application in KP’s FHR concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Next Generation Refrigerants for Domestic Heat Pump Water Heaters

To select next generation refrigerants and phase out R-134a, we conducted extensive drop-in tests of multiple new refrigerants. For laboratory investigations, we performed drop-in UEF (uniform energy factor) and FHR (first hour rating) tests on two heat pump water heaters (HPWHs), one uses a 220V compressor and 40-gallon water tank, and the other uses a 120V compressor and 50-gallon tank. The alternative refrigerants/blends include R-1234yf, R-1234ze, R-515B, R-513A, and R-516A, among which R-513A is A1, non-flammable and the others are in A2L category. All the alternative refrigerants required similar optimum system charges as the baseline R-134a. They resulted in similar or better UEFs than R-134a. R-516A led to best UEFs in both the HPWHs. The resultant discharge temperatures of the new refrigerants/blends are lower than R-134a, indicating good lubricant compatibility. The 120V HPWH delivered > 60 gallons first-hour rating using R-134a. The alternatives led to reduced FHRs due to their smaller volumetric vaporization capacities using the same compressor. Except R-515B, the other refrigerants/blends delivered FHRs > 60 gallons. R-516A produced better UEFs and FHRs in both the units than the other replacements. If considering flammability, R-513A would be the best option, which has similar UEF and FHR as R-134a.

Shen, Bo [ORNL] (ORCID:0000000336600393)↗
DOE OSTI2022

The corrosion effects of neutron activation of 2LiF-BeF 2 (FLiBe)

The redox effect of activation reactions in molten salts has been previously qualitatively defined but never quantified. Here, we quantify the corrosion effects of neutron activation of FLiBe. Rates of change of activation products are computed for FHR and ARC. The impact of activation reactions on salt chemistry is quantitatively compared to other environmental effects of similar consequences on the salt. The results are applied to a redox control case-study and an irradiation-corrosion experiment. The results are presented in a manner useful to the reactor chemist and designer concerned with corrosion and chemistry control and tritium management in irradiated FLiBe, and design of salt-irradiation experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Thermal hydraulics modeling of On-line refueling for the Advanced High Temperature Reactor (AHTR)

As part of the effort to move forward the Fluoride salt-cooled High-temperature Reactor (FHR) technology, the Advanced High Temperature Reactor (AHTR) is being developed at Oak Ridge National Laboratory (ORNL) and several other institutions. Due to its plank fuel design and inherently low heavy metal loading, which challenges fuel utilization, online refueling is considered as an option for improving the economic viability of this reactor. This paper presents the thermal hydraulic modeling of the online refueling for the AHTR. Analyses at the single channel and fuel assembly level have been performed in order to develop a model of the reactor. The reactor modeling has been integrated with CFD studies for characterizing the steady state conditions and the refueling transient. The simulations of the transient have been complemented by the analysis of the stability of the refueling operation. This work demonstrates the viability of online refueling from a thermal hydraulic standpoint and develops a modeling approach for this type of operational transient.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗