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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Yttrium Hydride Post-Irradiation Examination Plan

Micro nuclear reactors are expected to use metal hydride moderators to ensure thier compact design for the flexible operational purposes. Department of Energy’s Microreactor program therefore aims to provide infrasture and qualification bed for the development of micro nuclear reactors. Because micro nuclear reactors must be compact to offer a flexible operational purposes, metal hydrides are expected to be used as neutron moderator materials due to their high hydrogen number density. Los Alamos National Laboratory (LANL) directed by the DOE to provide a qualified fabrication techniques of metals hydride, especially the yttrium hydride (YHx) which provides advantages over other metal hydride candiates. Qualification of YHx involves fabrication, property, and performance pillars, which must be met for the deployment of microreactors. In the support of the qualification process metal hydride, LANL has developed fabrication techniques based in direct hydriding and powder compacting. For both manufacturing techniques, fresh material properties of metal hydrides have been reported in the literature. To establish the performance limits and the qualification parameters, an irradiation campaign has been initated at Idaho National Laboratory (INL). The irradation campaign aims to investigate the critical irradiated YHx properties as a function of neutron fluence and temperaute. The first phase of the irradiation campaign includes the effect of the operating temperature under neutron irradiation which provides initial data on the metal hydrides and informs the manufacturing processes. So that, drop-in irradations in Advanced Test Reactor (ATR) has been designed. This document depicts the post-irradation examination (PIE) plan of the ATR irradiated YHx. Objective of the PIE plan is to serve as a qualification basis for YHx as a neutron moderator material by understanding the impacts irradiation has on the physical and thermophysical properties; to establish a link between fabrication and the irradiated YHx properties; to institute a PIE strategy for other solid-state neutron moderator candidates, and; to inform micro nuclear reactor fuel performance and safety codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Materials for Small Nuclear Reactors and Micro Reactors, Including Space Reactors

Small nuclear reactors, including small modular reactors (SMRs) or reactors for space applications, rely on different materials than those typically applied in large-scale nuclear power plants. Examples include molten salts as cooling medium or fuel carrier, metal hydrides as high-temperature moderators, and fuels allowing for higher burnup. All of these also require novel structural materials, for which material interactions have to be understood. Fissionable and fissile materials, such as uranium or plutonium, are rarely considered in materials design other than for nuclear fuels. Similarly, the aspects of radiation damage, occurring during irradiation when a reactor operates, are unique to nuclear materials research. The handling of these materials puts further limitations on the materials science conducted for nuclear materials. All of these issues move research for these materials off the “main stream” of materials science, and cause it to be more easily conducted at national laboratories. However, nuclear reactors offer unique opportunities for carbon-neutral energy generation and have great potential to address if not solve problems arising from global warming. This special topic, sponsored by the TMS Nuclear Materials Committee, focuses on materials research for small nuclear reactors, both experimental and simulation/modeling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

High-Temperature Latent Heat Storage System using Transportable Heat Pipes for Versatile Integration with Emerging Microreactors

With the growing interest in the development of micro nuclear reactors and their application to microgrid, high-temperature thermal energy storage (TES) has also been receiving great attention as a technology to reinforce the energy utilization and profitability of this class of reactors. Given the importance and research needs of the TES technology, a research project is underway at Idaho National Laboratory (INL) to develop high-temperature TES that can be easily interfaced with emerging micro nuclear reactors. Currently under development is the Heat pipe-Integrated Thermal Battery (HITB), a novel latent heat storage design that enables flexible and robust integration with any design of currently considered micro nuclear reactors. The development of detailed design based on the proposed TES concept, HITB, requires not only development and validation of modeling and simulation capabilities, but also research that spans thermal performance analysis, component- and system-scale design optimization, and experimental demonstrations including high-temperature feasibility testing. The current research efforts (and plans) are divided into the following four areas: (i) Thermal and structural modeling and analysis at both system- and component-scales for evaluating performance and conducting design optimization, (ii) Small-scale experimental demonstrations for the key components of the proposed HITB design, (iii) Modelica-based dynamic model development for developing a control strategy under dynamic operation within integrated energy systems configuration, (iv) Integrated experimental testing and demonstration. This paper introduces the overall research efforts that have been and are being conducted to develop and demonstrate the novel design of the high-temperature latent heat storage device, HITB. The key design features, ongoing research efforts including the development of modeling and simulation capabilities, small-scale demonstrations of the HITB’s key functionalities prior to scaled-up experimental demonstration, are discussed. Then, this paper concludes with a remark on the future research plans for this project.

25 ENERGY STORAGE↗
DOE OSTI2022

Corrosion evaluation of metal foams in eutectic molten salts for high temperature latent heat energy storage application

High-temperature Thermal Energy Storage (TES) has drawn great attention as a technology that can increase the role and profitability of micro nuclear reactors in the decentralized clean energy market. Various research needs have been proposed by Idaho National Laboratory (INL) in "the Integrated Energy Systems: 2020 Roadmap" with the goal of utilizing the high temperature of 550 degrees or more generated in the 4th generation reactor to industrial purposes from the standpoint of integrated energy systems [1]. With the recent development of the 4th generation nuclear power plant, including micro nuclear reactor, there is a great deal of interest in researching how to efficiently couple a heat source to an industrial process through thermal processing. Texas A&M University (TAMU) and INL are currently collaborating to develop a novel latent heat storage design called HITB (Heat pipe-Integrated Thermal Battery). As part of the small-scale experimental demonstration research for HITB, a study for storage medium selection was conducted at TAMU. Eutectic salts have attracted interest as high-temperature heat storage medium for HITB. Especially, chloride molten salts and fluoride molten salts are considered promising candidates due to their high melting temperature, thermal stability, and high latent heat of fusion. Since eutectic salts have poor thermal conductivity in general, it is critical to distribute materials with high thermal conductivity evenly to facilitate the charging and discharging of heat. In this study, metal foam was considered in the HITB system to overcome the poor heat transfer characteristics of eutectic salts. In order to improve heat transfer, it is necessary to study materials with high thermal conductivity, corrosion resistant to molten salts, such as fins, extended surfaces, particles, microcapsulation, or foam structure. Porous structure of copper or aluminum can increase heat transfer area, form a thermal transfer network, and increase the effective thermal conductivity of thermal storage medium [2]. Since the melting temperature is required to be at least 450? in the current HITB design, various eutectic salts are being considered. FLiNaK and FLiBe, which are famous for fluoride salts, are good candidates, but due to the sharp rise in the price of LiF recently, they are excluded for economic reasons to be applied to large-capacity thermal energy storage. On the other hand, chloride salts are very cheap and easy to obtain, have a high melting temperature, and have a high latent heat of fusion, so they are recently in the spotlight as a phase change material. However, because of the hygroscopic nature of chloride salts, HCl gas due to impurity is easy to be released and is particularly vulnerable to corrosion. Chloride eutectic salts have been tested and suggested for metal alloys that are particularly resistant to corrosion, such as SS304, SS316, Hastelloy, Inconel 625, Incoloy 800H, which has low thermal conductivity, and is very expensive [3-6]. Porous materials with high thermal conductivity such as copper and aluminum have been tested for stability in fluids such as paraffin or water as metal foam or metal fin structures, but high temperature corrosion tests were not performed on various eutectic salts yet. Therefore, this study investigated the high temperature corrosion characteristics of metal alloys, such as C10100 foam, C10100 plate, 6101 alloy foam, and SS316 plate, while immersed in the candidate eutectic salts. A total of 20 tests for SS316 coupons and 10 tests for C10100 coupons were performed to ensure the repeatability of the present measurements. Since it is difficult to completely remove the salt inside metal foam due to the complex inner structure of the metal foam, only the surface condition was observed with a microscopy and scanning electron microscopy (SEM) image, except for measuring the corrosion rate and average mass loss.

25 ENERGY STORAGE↗
DOE OSTI2022

A CONDUCTION-BASED HEAT PIPE MODEL FOR ANALYZING THE ENTIRE PROCESS OF LIQUID-METAL HEAT PIPE STARTUP

The Heat Pipe-cooled Microreactor (HPM) is one of the micro nuclear reactor designs under active study at the U.S. Idaho National Laboratory. Among the major concerns of HPM research is to understand the startup behavior of the heat pipe-cooled system associated with the startup of the liquid-metal heat pipes initialing from frozen state. The startup of liquid-metal heat pipes typically involves a number of nonlinear mass and heat transport processes including the phase change from solid to liquid and vapor. Hence, it is still a huge challenge to simulate the liquid-metal heat pipe startup using conventional CFD methods and software. The major difficulties of numerical CFD modeling come from the phase-change process, multiphase interaction, microporous wick flow, and compressible gas dynamics that occur during startup of the liquid-metal heat pipes. This paper proposes a simplified conduction-based method to provide practical insights into the entire startup process of the liquid-metal heat pipes while mitigating the challenges of addressing all the complex physics. We discuss the theoretical basis and modeling assumptions to analyze the liquid-metal heat pipe startup from frozen state based solely on heat-conduction equations. Then, the proposed model is implemented into the commercial CFD software to verify the model performance. The model prediction results are discussed via the comparison with the experimental data obtained from sodium heat-pipe startup experiments.

42 ENGINEERING↗
DOE OSTI2020

Technical and Licensing Considerations for Micro-Reactors

The U.S. Nuclear Regulatory Commission (NRC) has interacted with vendors pursuing the commercialization of micro-reactors (i.e., reactors capable of producing about 1 MW(th) to 20 MW(th) of energy from nuclear fission). It is envisioned that micro-reactors could be assembled and fueled in a factory and shipped to a site. Many of the sites are expected to be remote locations requiring off-grid power or in some cases military bases. The objective of this effort is to explore the technical issues and the approach required to reach a finding of "reasonable assurance of public health and safety" for this new and different class of reactors. The analysis performed here leverages available micro-reactor design and testing data available from national laboratory experience as well as commercial design information to explore technical issues. Some factors considered include source term, accidents that would need to be analyzed, and the extent of the probabilistic risk assessment (PRA). The technical evaluation was performed within the framework of the Licensing Modernization Project (LMP) to identify licensing basis events, classification of structures, systems and components, and defense-in-depth needed to provide regulatory certainty. With this framework and technical evaluation in mind, the scope and content of a micro-reactor licensing application is discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Powering Data Centers with Clean Energy: A Techno-Economic Case Study of Nuclear and Renewable Energy Dependability

Rising data demands from artificial intelligence (AI) and large language models (LLMs) generating images, videos, and text have prompted increased need for larger and more robust data centers in the United States. Major companies interested in these larger data centers face the choice of linking them to existing regional grids, building stand-alone power supplies onsite, or a combination of both. The request, review, and approval process for new transmission lines to grids in the United States, however, has grown in recent years to times spans rivaling those of new construction for nuclear power plants. Building an islanded power supply for each data center is therefore becoming a prominent option. In this case study, several technologies are modeled in techno-economic simulations for long-term system costs subject to fixed electricity demand from a singular data center. A 250 MWe data center is assumed with additional 50 MWe for resiliency. Techno-economic simulations are conducted using the Holistic Energy Resource Optimization Network (HERON) software, which is a part of the Framework for Optimization of Resources and Economics (FORCE) tool suite. Technologies considered include solar, wind, lithium-ion batteries, and several types of nuclear reactors: large-scale reactors, small modular reactors, and microreactors. A low- and high-cost estimate for each technology is assumed to develop a range of expected economic performance. Low-cost estimates included several clean energy production tax credits. Different combinations of renewable energy generators with nuclear reactors are considered, ranging from a fully renewable-powered data center to a fully nuclear-powered data center. Historic time series of wind and solar availability from the Texas grid are used to train a reduced order model; this model then generates unique time series with similar characteristics of the training dataset. Multiple scenarios of weather and subsequent operations are simulated for each renewable-nuclear combination to determine total costs throughout the project lifetime. Fully renewable-powered configurations required large amounts of installed capacity (GW scale) in the simulations to meet the fixed demand of the data center. This is due to some scenarios in the historical dataset which captured low-wind and low-solar days, requiring over-building of these technologies as well as batteries to compensate for the low amounts of electricity generation. Fully nuclear-powered configurations outperformed the fully renewable and mixed renewable-nuclear configurations in terms of cost, with ranges between $1B and $10B in 2023 USDs compared to $40B+ for fully renewable configurations. Of the nuclear technologies, small modular reactors performed better economically than large-scale nuclear models due to lower projected capital costs, and both performed better than the microreactor models. These results demonstrate the applicability of firm, dispatchable electricity resources from baseload generators like nuclear power plants for operating facilities that run at constant power without daily variability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Fate and transport of unruptured tri-structural isotropic (TRISO) fuel particles in the event of environmental release for advanced and micro reactor applications

Advanced nuclear reactor designs and advanced fuel types offer safety features that may reduce environmental consequences in an accident scenario when compared to conventional reactors and fuels. One advanced reactor fuel is tri-structural isotropic (TRISO) fuel particles which are approximately 0.9 mm in diameter. TRISO particle mobility, assuming the particle is unruptured and the encapsulated radionuclides are contained, was explored through a theoretical examination of mobility through atmospheric, soil and groundwater, surface water, and non-human biota transport pathways. TRISO particles are too large and dense to travel in the atmosphere except under extreme conditions. TRISO particles are too large to penetrate most soil profiles and so cannot be transported with groundwater. TRISO particles will settle out of the water column in surface waters but the transport will depend on the energy of the body of water and likelihood of extreme dispersion events. TRISO particles could be transported by non-human biota. The size of TRISO particles could allow them to be intentionally moved by non-human biota if they are ingested as a gastrolith or if they are mistaken for something in the organism’s diet. Generally, TRISO particles will have reduced environmental mobility compared to releases of radionuclides in the event of an accident in a conventional nuclear reactor. The extent of transport has implications in emergency planning zone designations and other considerations for licensing and deploying TRISO-fueled reactors. Finally, further research and experimental work exploring TRISO particle mobility is required to understand the full environmental mobility of TRISO particles.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Neutron moderation modules

Provided herein is a neutron moderation module and a thermal-neutron nuclear micro-reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Equilibrium Core Model for Micro Pebble Bed Reactors Using OpenMC

Estimating the equilibrium state for pebble bed reactors (PBRs) presents complex challenges as it requires simultaneous consideration of changes in the pebbles’ movement as well as their fuel compositions. Whereas traditional approaches use multigroup diffusion codes for neutronics calculations of PBRs’ equilibrium state, the double-heterogeneity of PBRs complicates neutron cross-section generation. Continuous-energy Monte Carlo (MC) methods are better suited for detailed PBR analysis because of their natural handling of double-heterogeneity, but they demand substantially more computational resources. Here, this study introduces a novel method for efficiently estimating the equilibrium state in small and micro PBRs with reduced computational cost. The method is anticipated to accelerate the processes of core design and performing parametric studies for utilizing advanced fuel and structural materials. The HTR-10 reactor design was used for validating the method’s predictions and evaluating its computational efficiency. When compared to reference calculation values from the literature, criticality (k-effective) was predicted to be approximately within the margin of error of the MC transport calculation, average core power density (in megawatts per cubic meter) was predicted within 2.5% relative error, and maximum thermal flux (10 13 n/cm 2 .s −1 ) was predicted within 1.8% relative error. The calculated inventory of fission products and fuel composition in the equilibrium core were within 15% and 16.6%, respectively, when compared to reported values from the literature. The difference is attributed to variance in the considered values of the core temperature, which was found to significantly affect the depletion analyses.

Equilibrium core↗
DOE OSTI2022

Design optimization of gap distance for the capillary limitation of a heat pipe with annular-type wick structure

In this study, an experimental investigation was conducted on the rising height and contact angle of fluid in an annular wick-type heat pipe. The annular wick-type heat pipe was characterized by a small gap between the wick structure and tube wall, which compensated for the pressure drop along the porous media and created additional capillary force. To describe and model the advantage of this gap, the rising of a wetting liquid in the gap between a vertical solid plate and a mesh (with a small angle between them) was experimentally measured and analyzed. An additional experiment was performed to investigate the effect of curvature on the capillary rise using tubes and meshes of varying radii. Resultantly, we confirmed that the linear combination of the contact angles of the solid plate and mesh could be applied to calculate the rising height from the Laplace–Young equation. Furthermore, the effect of curvature on the rising height of the liquid was negligible. These results were extended to the investigation of finding the optimal gap distance for the annular wick-type heat pipe by referring to previous studies. We observed that a gap distance of 1.27 mm provided the largest permeability (K) over the effective pore radius (r eff ) value for a heat pipe with ethanol, which in turn resulted in the highest capillary limitation. For a sodium heat pipe, a gap distance of 0.84 mm resulted in the highest capillary limitation.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2020

Calculation of Friction Factors in Heat Pipes Using CFD in Support of the Sockeye Code

A heat pipe is two-phase heat transfer device which relies on surface tension and capillary pressure to provide a very efficient heat transfer mechanism. Currently, there is interest in designing micro nuclear power reactors using high temperature heat pipes to provide totally passive cooling. To support this, Sockeye is being developed as a heat pipe simulation capability. To support the development of Sockeye, individual phenomena must be tested. This work focuses on applying a CFD model to evaluate friction factors for use in Sockeye. A CFD model has been developed in Nek5000 which simulates each component fluid, i.e. sodium liquid and vapor, independently and captures the effects of the wick on friction factor. The model tests the effect of the wick permeability and the thickness of the wick occupied by each fluid component. The thickness of the wick occupied by each component can be directly related to the local vapor volume fraction from the Sockeye heat pipe model. Results from the CFD model indicate that friction factor in the liquid flow decreases with increasing wick permeability, while the friction factor in the vapor flow increases with increasing wick permeability. In the liquid, this is likely a direct result of the increased cross sectional area available for the flow. For the vapor, this is explained as the wick acting like a rough wall, while the increased area has only a negligible impact. Results for the tests of the thickness of the wick occupied by each fluid showed a thresholding behavior for both liquid and vapor. When the thickness of the wick occupied by each fluid was increased, it increased the effect of the wick, but only up to a certain point. Once a certain thickness was reached, the friction factors remained constant. When the fluid occupied only a very small fraction of the wick, friction factors returned to nominal values.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Proceedings of the international conference on physics of reactors - Physor 2022

The theme for this year's conference is 'Making Virtual a Reality: Advancements in Reactor Physics to Leap Forward Reactor Operation and Deployment'. As we leap forward to the future of nuclear technology, micro-reactors, small modular reactors, advanced reactors, and high energy LWR fuel are at the forefront of the topics.The Physor 2022 will cover a broad range of topics in thirty standard and special sessions tracks, and ten panel sessions gathering more than 350 papers.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Regulatory Review of Micro-Reactors: Initial Considerations

The U.S. Nuclear Regulatory Commission (NRC) is working to have an effective and efficient mission readiness for reactors that differ considerably from those currently licensed. Micro-reactors, that is, reactors that have a power of roughly no more than 10 MW, are one class of these advanced reactors. This report is one of the first steps taken to articulate the many technical and regulatory issues that will need to be addressed for NRC to have the ability to review license applications for micro-reactors. Many of the issues center around the fact that a) these reactors may be operated remotely or semi-autonomously and b) it will be difficult to analyze risk from new, unique, technologies. Initial thoughts are given on how probabilistic methods could be used to determine risk and how the current approach for reviewing non-power reactors could be useful for micro-reactors.

42 ENGINEERING↗
DOE OSTI2024

Preliminary Analysis of a Nuclear Grade Sandwich Composite for Transportable Micro-Reactor Shielding

Transportable microreactors pose a unique shielding challenge compared with conventional reactors. Many vendors are seeking to transport a microreactor to a site, operate it for a number of years, and then transport it to a secondary site. This involves ensuring the reactor has sufficient shielding during operations and post-irradiation transportation to ensure the safety of those around it, all while ensuring it can be transported via traditional means. This work examines the use of a nuclear grade sandwich composite (NGSC) which seeks to combine the reactor pressure vessel and biological shielding into a single unit. Through this work, it was determined that the NGSC was sufficient to adequately reduce neutron and gamma current exiting the structure to warrant further investigation as a potential shielding option for microreactors.

42 ENGINEERING↗
DOE OSTI2023

Micro-reactor core mechanical support

A nuclear reactor core mechanical support bracket is disclosed. The support bracket includes a housing, a spring disposed internally within the housing, a shaft slidingly disposed within the housing, a shaft travel pin, and a flange. The shaft is configured to engage the spring to compress and decompress the spring as the shaft travels in and out of the housing. The shaft travel pin controls the travel of the shaft. The flange is configured to mount the nuclear reactor core mechanical support bracket to a canister of a nuclear reactor. The shaft includes an inset configured to interface with a nuclear reactor core component.

Aleshin, Yuriy↗
DOE OSTI2026

Effects of Uranium Impurities in Downblended HEU on HTGR Performance

Many advanced reactor designs require fuel enriched between 5% and 20% 235 U. To assist in producing fuel at these enrichment levels, government-owned inventories of highly enriched uranium can be downblended. However, fuel produced from these inventories contain uranium impurities that are not often found when enriching natural uranium or accounted for when modeling reactor cores. To address this concern, this work models reactor designs like the X-energy Xe-100 and the Ultra Safe Nuclear Company’s Micro Modular Reactor, and compares their performance with fuel from enriching natural uranium to fuel from downblended highly enriched uranium. This paper evaluates the models based on the effective neutron multiplication factor, k eff , effective delayed neutron fraction, β eff , and energy- and spatially dependent neutron flux, ϕ, as well as the fuel, coolant, moderator, and total reactivity temperature feedback coefficients, α F , α C , α M , and α T . The results show that the fuel from downblended highly enriched uranium inventories leads to differences in each of the metrics, especially in the keff values. In the Xe-100–like and Micro Modular Reactor–like models, k eff changes by about 1400 pcm and up to 1200 pcm, respectively. Total reactivity feedback coefficients α T are negative with the impure fuels and the keff values remain above 1 for each core configuration and fuel composition. These results show that the impure fuel compositions do not necessarily prevent achieving key design parameters, such as cycle length, or from operating in a safe condition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Advanced Reactors Development in USA

Small and micro-reactor advanced reactor development in the USA. Advanced Reactor programs, nuclear renaissance, and INL as a test bed facility. These slides are for presentations in general on the past, INL and nuclear engineering worldwide, GEN IV reactor designs, SMRs and micro Rx designs, multiple commercial designs, and what INL as a test bed means and who uses this to their advantage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗