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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 OSTI2022

A Low-Enriched Uranium (LEU) option for the conversion of FRM II

The Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRM II) is Germany’s most powerful research reactor and uses Highly Enriched Uranium (HEU) fuel enriched at 93 %. The Technical University of Munich (TUM) operates the reactor and has been mandated to convert the FRM II to a lower-enrichment fuel. In preparation for the lower-enrichment fuel downselection, planned for 2023, TUM is evaluating multiple conversion scenarios. In this paper, it is demonstrated that a conversion of FRM II to Low-Enriched Uranium (LEU) is scientifically possible when using the novel monolithic U-10Mo fuel system, which is currently in the qualification process. As a research reactor, safety criteria, neutron flux distribution and cycle length are key metrics. To begin, neutronic and thermal-hydraulic models were created based on current technical drawings and relevant material properties. Specifically tailored computational methods and coupling schemes have also been developed to properly evaluate all relevant characteristics for the conversion of the FRM II reactor. Using these models and methods, a systematic parameter study was performed to explore the defined design space. The results of that study indicate that many LEU designs using monolithic U-10Mo appear to be viable to convert FRM II to LEU. One example of an LEU solution for the FRM II reactor that fulfills the stated safety, compatibility and scientific performance requirements is discussed in more detail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Unveiling damage mechanisms of chromium-coated zirconium-based fuel claddings at LWR operating temperature by in-situ digital image correlation

Here, we investigate the coupled thermomechanical fracture mechanisms of coated nuclear fuel claddings at Light Water Reactor (LWR) operating temperatures. These coated claddings are a highly attractive, near-term solution, which addresses the demands for accident-tolerant fuel systems and provide greater oxidation resistance. However, the fracture mechanisms of these coatings, which may create channels for oxidation ingression, must be fully understood prior to implementation. Thus, high-temperature expanding plug experiments were conducted on coated cladding specimens at a temperature of 315 °C, consistent with the operating environment of LWRs. Furthermore, in-situ thermomechanical deformation was measured with stereo digital image correlation during heating and mechanical testing to separately resolve contributions of thermal and mechanical strain. Digital image correlation, supported by acoustic emissions (AE) detection, was also leveraged to track cracking activity during loading. Coating fracture was found to initiate at total hoop strains of 0.34%. The thermal deformation of the coated claddings was investigated via finite element simulations, revealing a bidirectional stress-state within the coating with axial and circumferential strains of 0.026 and 0.031%. This bidirectional stress-state was attributed with the generation of off-axis fracture pattern within the coating as identified via post-experiment scanning electron microscopy. Thus, this study unveiled critical, coupled thermomechanical mechanisms governing the coating fracture of coated claddings at LWR temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Developing Materials for Extreme Environments and Their Use in Nuclear Waste Canisters and Reactors - 20559

The paper focuses on the development of materials which can be utilized in extreme industrial environments. The paper discusses the advancement of materials and their application in components which are utilized in extreme industrial environments as such that of the nuclear field and especially in the nuclear waste field. Nuclear power plants have become an indispensable part of the electricity generation, medical, food processing and many other industries. Though burning small amount of fuel as compared to conventional power plants, they also create waste which is not easy to handle as other types of wastes. The nuclear waste sitting at the bottom of spent fuel tanks around the globe has raised public concerns on how it will be dealt with. Europe has started to give an answer by working on geological repositories. Though till yet, waste has been cooled in spent fuel tanks and after processing, is precisely placed in nuclear waste canisters and casks and transported to sites across the country for further cooling and storage. Though in all this process there had been a great deal of advancement on the type of casks and canisters used and the processes for transporting nuclear material and waste. Often reports about radioactive material leakage, moisture content inside barrels etc. has surfaced. Be it official or unofficial report, there seems a technical failure on how the nuclear waste has been stored and placed at facilities around the globe. The other aspect of this material development is to be beneficial for nuclear reactors. As the technology gets advanced, nuclear reactors change their titles from Gen II to Gen III and so on. The new fleet of upcoming reactors will be exceptionally safer, reliable and more efficient. But these benefits will also bring drawbacks with them. As these reactors will operate at high temperature (above 800 degree Celsius) they also ask for advancement in materials used in the reactor core, fuel used, reactor control etc. The paper tries to answer such concerning questions. The work shows the study done on different material that can be better candidates for extreme environments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Thermal Integration of Advanced Nuclear Reactors with a Reference Refinery, Methanol Synthesis, and a Wood Pulp Plant (Rev.1)

The present report is intended to provide process flow diagrams (PFDs) and energy and mass balance data sheets for a U.S. industrial sector subset with which nuclear heat and power could be integrated—a subset that includes the oil refining, methanol and pulp and paper industries. Coupling options for integrating nuclear energy into these industries are quantitatively outlined for reference systems, and future work will extend this analysis in greater detail. Opportunities for integrating small modular nuclear reactors (SMNRs) were investigated for each of the industrial process configurations. Aspen HYSYS and Cycle-Tempo models for a high-temperature gas-cooled reactor were developed to evaluate the proposed integration. This introductory evaluation provides a general description and assessment of the operating principles, reactor coolant core outlet temperature, and reactor size to be integrated with industry. The industrial processes of oil refining and the production of methanol, pulp and paper were simulated by using Aspen HYSYS, Aspen Plus, and the PRELIM (Petroleum Refinery Life Cycle Inventory Model) tool to develop process details. Cycle-Tempo models then extend the process modeling results to obtain net energy demands (e.g., heat, steam, and electricity) when accounting for process steam and waste heat recovery. This information is intended to foster the analysis of integrating an SMNR to decarbonize industrial facilities. The SMNR would provide reliable, competitive, and sustainable clean energy while reducing carbon emissions and other environmental impacts, such as water withdrawals, consumption, and contamination. The refining industry, exhibited in Figure ES1, is a leading consumer of fossil -fuel-based heat, power, and hydrogen in the U.S. industrial sector, generating over 164 million metric tons (MMT) of CO 2 emissions in 2023. The overall mass and energy pertaining to a generalized complex refinery in the United States is reflected in Figure ES1, along with energy metrics regarding integration with a nuclear power plant (NPP). Data sheets were developed to indicate the energy requirements for the overall refinery and each refinery process. The data sheet for the overall refinery is shown in Table ES2.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Modeling for a Digital Twin-Based Remote Operation System Framework

New reactor designs and technologies are being developed to grow and advance the nuclear industry. Microreactors are one of the many new concepts for advancing the industry. Microreactors are very small reactors generally designed to have an operating power of 20 MWth or less. They are ideal for many applications in which it would not be feasible to have a large-scale reactor, such as powering remote communities, military bases, and mining sites. Many of these applications currently rely on diesel generators for power, and replacing those generators with the carbon-free energy of a microreactor is a major driving factor for microreactor development. However, most of the use cases for microreactors are in isolated locations where construction and labor costs are much higher. Microreactors will need to be comparable to other energy production methods for their deployment to be successful. Remotely operating the microreactor has a great potential to benefit the economics and make it more cost competitive. With remote operations, the operation facility location could be strategically chosen based on factors such as construction costs, workforce size, etc. The benefits of remote operations could be leveraged even more if the remote operation system is semi-autonomous. If the remote operation system is semi-autonomous, more microreactors could be operated and monitored from one remote operation facility. Additionally, with the system handling some tasks for the human operator, it could reduce the number of operating staff necessary for the microreactor. Digital twins can be used to introduce a level of automation to the remote operation system. Digital twins are capable of component monitoring, system operation and control, and predictive performance [1]. All of those features are important for a successful semi-autonomous remote operation system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Dispatch analysis of flexible power operation with multi-unit small modular reactors

The relevance of nuclear power plant flexible power operation (FPO) is rising due to increased penetration of variable renewables. Small Modular Reactors (SMRs) such as NuScale are envisaged to have multiple units on one site. This provides opportunities for coordinating FPO between units, but also introduces challenges as common services (e.g., refueling equipment) must be shared between units and fuel loadings may be standardized. It is therefore important to quantify whether FPO with SMRs is economically beneficial. Here, this paper quantifies the economics of FPO on different timescales to analyze refueling outages of such multi-unit SMRs. A 24-h price-taker profit maximization of SMR operation is first solved considering revenues from wholesale power and ancillary service (AS) markets. The results are then used to find the most profitable operation strategy over several years for multiple units at the site accounting for the physics and materials limits on NPP operation, including refueling outages. Results show a small but appreciable participation of nuclear into AS, contributing 4% of revenues. Due to the longer fuel cycle, over a decade of operation the units’ refueling outages drifted by 3 months, ultimately leading to refueling during summer.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Device and method for improving specific energy requirement of plasma pyrolyzing or reforming systems

Devices and methods for reducing the specific energy required to reform or pyrolyze reactants in plasmas operating at high flow rates and high pressures are presented. These systems and methods include 1) introducing electrons and/or easily ionized materials to a plasma reactor, 2) increasing turbulence and swirl velocity of the flows of feed gases to have improved mixing in a plasma reactor, and 3) reducing slippage from a plasma reactor system. Such plasma systems may allow plasma reactors to operate at lower temperatures, higher pressure, with improved plasma ignition, increased throughput and improved energy efficiency. In preferred embodiments, the plasma reactors are used to produce hydrogen and carbon monoxide, hydrogen and carbon, or carbon monoxide through reforming and pyrolysis reactions. Preferred feedstocks include methane, carbon dioxide, and other hydrocarbons.

Bromberg, Leslie↗
DOE OSTI2022

Overview of Base Model in Parametric Studies Specific to Performance of U-Mo Plates

This paper provides an overview of the base model specifically developed to perform parametric sensitivity studies on the U-10Mo monolithic fuel system. U-Mo monolithic fuels are being considered for the conversion of test reactors into high-performance research reactors that operate using proliferation-resistant, low-enriched uranium (LEU) fuels. These plate-type fuels contain a high-density, low-enrichment fuel sandwiched between zirconium diffusion barriers and encapsulated in aluminum claddings. All U.S. high-performance research reactors have released the designs of their LEU monolithic fuel reactor cores. These designs include nearly 50 distinct fuel plate geometries with different operational parameters. Consequently, a single generic plate geometry representing all the extreme points in this design matrix is unrealistic. To evaluate the performance for various parameters, a set of sensitivity studies was performed. These studies considered various input parameters (i.e., geometric, operational, and material property-related). The results revealed valuable information about plate performance and the sensitivity of this performance to various modeling inputs. To establish a reference state for comparing these result, base model featuring representative irradiation conditions was developed. To capture in-reactor behavior accurately, incorporation of representative constitutive models capable of evolving properties with respect to temperature, irradiation time, and burnup was needed. The behavioral models considered burnup-dependent properties, swelling, creep, and degradation. This paper introduces the base model created for the parametric sensitivity studies. The detailed description of the procedure includes the model geometry, model discretization, thermo-mechanical coupling, material properties and behavioral models. This paper also provides selected results and assesses the performance of the base model.

42 ENGINEERING↗
DOE OSTI2025

Development and Demonstration of a Prototype Molten Salt Sampling System

Molten salt reactors (MSRs) offer potential operability and safety advantages when compared to commercial light water reactors (LWRs). However, operating experience with MSRs is sparse in comparison to what exists for LWRs. Further, the chemical and isotopic composition of the fuel and/or coolant salt is dynamic and difficult to characterize continuously, posing potential safety, operability, and safeguards unknowns that need to be addressed. A molten salt sampling system (MSSS) is regarded as a necessary subsystem within first generation MSRs used to obtain samples of salt for chemical and isotopic analysis in support of the need to monitor and control salt composition during operation. The MSSS is being developed using the Safety-in-Design (SiD) methodology, which incorporates incremental integration of safety analysis into the design process. The MSSS conceptual design emerging from the application of the early stages of the SiD methodology consists of a sample collection system and its housing, a freeze port, and inert gas control and delivery systems. This article describes the prototypes developed to test the functions of these MSSS subsystems, presents the results of testing in both dry and molten salt environments (including reliability data collection performed in accordance with the principles of SiD and the development of a semiquantitative fault tree model), and summarizes the opportunities for future design and testing enhancements based on the results of prototype testing.

molten salt reactor↗
DOE OSTI2023

Post Irradiation Examination of Pressurized Water Reactor Stainless Steel Internal Components

Internal structural components of pressurized water reactors (PWR), such as baffle former bolts, are subjected to significant neutron irradiation and mechanical stresses at elevated temperatures during plant operation. Over the long operation of the power plant, these conditions lead to potential degradation and reduced load-carrying capacity of these bolts. To understand property degradation more fully, to confirm the results of experimental irradiation programs, and to predict operational lifetime performance of structural materials in internal components, post irradiation examination of harvested materials from operating nuclear reactors is required. In this work, two high fluence 316 stainless steel baffle former bolts were retrieved from a commercial Westinghouse two-loop downflow type PWR and then sectioned for analysis via mechanical testing and microstructural characterization. The irradiation damage fluctuated along the bolt length with damage levels from 15 to 41 displacements per atom with the bolt head receiving approximately twice the neutron damage levels as that of the bolt thread section. Mechanical testing evaluations showed extensive irradiation hardening and a sharp decrease in fracture toughness in all parts of the bolt, though with limited variation along the bolt length. Microstructural characterization using analytical scanning transmission electron microscopy and atom probe tomography showed significant radiation induced precipitation, segregation, and dislocation loop and cavity formation. However, unlike the mechanical behavior, there was considerable variation along the bolt length but opposite of what may be expected, with more precipitation and cavity formation in the bolt thread where the neutron radiation dose is less than in the bolt head where the neutron radiation dose is higher. The cause of this variation is likely due to gradients in temperature, neutron energy spectra, and gamma irradiation.

Lach, Tim↗
DOE OSTI2024

Opportunities, Challenges, and Research Needs for Remote Microreactor Operations

As the nuclear industry develops new advanced reactor technologies, many companies are embracing this advancement by pursuing the development of microreactors. The term microreactor generally refers to a nuclear reactor with an operating power of 20 MW(thermal) or less. The power range of microreactors makes them appealing for many use cases, such as powering remote communities, mining sites, and military bases. Most of the microreactor designs being pursued are expected to incorporate remote facility operations into the final product. However, no framework has yet been developed to determine what remote operations systems require for reliable, resilient, and secure operation of a microreactor. Here, this work identifies the research needs for challenges that are unique to remote operations and monitoring for microreactors, specifically regarding instrumentation and control, communication methods, regulatory requirements, and operational policies. The types of commands and sensor measurements that must be transmitted between the facilities, as well as methods for verifying the trustworthiness of these signals, are assessed. This work evaluates the security, reliability, and performance requirements that must be met when considering the selection of communication hardware and protocols for use in remote operations. Also, an assessment was performed to study how remote operations fit within current regulatory requirements and what may need to be updated in regulatory policy to allow for remote operation. Finally, the operational contingencies unique to remote operations that must be in place for responses to abnormal events are identified. This paper identifies the challenges and research opportunities within the areas of importance for the design of remote operation systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Modeling and Simulation of Fuel Burnup in Pebble Bed Reactors

Modeling and simulation of fuel burnup plays important roles in reactor design, operation, safety, and security as well as nuclear material control and accounting (MC&A) [1]. This task is uniquely challenging for pebble bed reactors (PBR) because the pebbles are continuously added and recycled into the reactor core, and their paths through the core are random. To address this problem, we present two simulation models in this paper. Brookhaven National Laboratory (BNL) developed a simple lattice model of a PBR in Serpent software to generate used pebble isotopic concentrations. The benefit of using Serpent software in this specific application is that it helps streamline the data generation process without having to use too many independent software codes in combination to achieve a simple task. For example, transport, burnup and zero power decay can be implemented in a single pass. Three-dimensional core models were developed using Serpent to simulate the burnup process of 5 subject pebbles starting from fresh till they reach nearly target burnup, with each pebble placed in one of the five artificially designated radial channels to capture the changing neutron spectra along the core radius. Equilibrium isotopic concentrations were assumed in all other pebbles in the core. To provide a verification for the Serpent isotope transmutation and decay results, Oak Ridge National Laboratory (ORNL) performed simple SCALE/ORIGEN calculations using the average neutron spectra calculated by Serpent for each of the 5 pebbles. The 252-group neutron spectra from Serpent were then used by ORIGEN to produce the one-group library for depletion and decay calculations. The isotopic concentrations of a few nuclides of interest and neutron and gamma source terms produced from the ORIGEN calculations were compared with the ones from the Serpent calculations. The model simulated in this work was based on the Pebble Bed Modular Reactor (PBMR)-400 design because many data needed for the simulation such as core power profiles, fuel and reflector temperatures, and equilibrium core composition are publicly available. In this paper, we will compare the results between these two approaches and benchmark the results against a set of well-established simulation results for PBMR-400.

Dim, Odera↗
DOE OSTI2023

Modeling and Simulation of Fuel Burnup in Pebble Bed Reactors

Modeling and simulation of fuel burnup plays important roles in reactor design, operation, safety, and security as well as nuclear material control and accounting (MC&A) [1]. This task is uniquely challenging for pebble bed reactors (PBR) because the pebbles are continuously added and recycled into the reactor core, and their paths through the core are random. To address this problem, we present two simulation models in this paper. Brookhaven National Laboratory (BNL) developed a simple lattice model of a PBR in Serpent software to generate used pebble isotopic concentrations. The benefit of using Serpent software in this specific application is that it helps streamline the data generation process without having to use too many independent software codes in combination to achieve a simple task. For example, transport, burnup and zero power decay can be implemented in a single pass. Three-dimensional core models were developed using Serpent to simulate the burnup process of five subject pebbles starting from fresh till they reach nearly target burnup, with each pebble placed in one of the five artificially designated radial channels to capture the changing neutron spectra along the core radius. Equilibrium isotopic concentrations were assumed in all other pebbles in the core. To provide a verification for the Serpent isotope transmutation and decay results, Oak Ridge National Laboratory (ORNL) performed simple SCALE/ORIGEN calculations using the average neutron spectra calculated by Serpent for each of the five pebbles. The 252-group neutron spectra from Serpent were then used by ORIGEN to produce the one-group library for depletion and decay calculations. The isotopic concentrations of a few nuclides of interest and neutron and gamma source terms produced from the ORIGEN calculations were compared with the ones from the Serpent calculations. The model simulated in this work was based on the Pebble Bed Modular Reactor (PBMR)-400 design because many data needed for the simulation such as core power profiles, fuel and reflector temperatures, and equilibrium core composition are publicly available. In this paper, we will compare the results between these two approaches and benchmark the results against a set of well-established simulation results for PBMR-400.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2022

Mechanical response and microstructure evolution from multiaxial stress relaxation in textured zircaloy nuclear cladding

Modern energy markets provide motivation for nuclear reactors to increase operational flexibility and load following capacity of existing reactor designs. The operational power changes result in fuel pellet expansion and contraction resulting in pellet-cladding mechanical interaction (PCMI) and high cladding stresses that are strongly related to stressed corrosion cracking (SCC) which can lead to fuel rod failure. This work aims to quantify the stress fields in cladding with a focus on multiaxial stress relaxation response from an imposed strain and the resulting microstructure changes in textured zircaloy claddings. Stress relaxation tests resulting in multiaxial stress fields are performed to measure the material response of Zircaloy-4 nuclear cladding. Multiple experiments using samples from the same piece of cladding are performed so samples with the same processing history may be removed at strategic points throughout mechanical testing for microstructure characterization. The microstructure evolution is analyzed with mechanical response to evaluate the static recovery impact on crystallographic texture and grain morphology. Experiments are performed at Idaho National Laboratory using an experimental device capable of applying internal pressure and independent axial load to commercial nuclear cladding at elevated temperatures with stress- or strain-controlled experimental capability. X-ray diffraction and electron backscatter diffraction are used for microstructure analysis. Rheological models are used to visualize the multiaxial stress-strain relationships and isolate elastic, plastic, and viscoplastic material properties for analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Buffer-IPyC separation process in TRISO fuel particles simulated with Bison code

During High Temperature Gas-cooled Reactor (HTGR) operation, tristructural isotropic (TRISO) coated-particle fuel undergoes irradiation-induced changes in morphology and thermomechanical properties. Experimental results from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program show, among other things, the mechanism of gap formation between the buffer and inner pyrolytic carbon (IPyC) layers, which could be explored further via computational simulations using the Bison code. Two simulation models were developed, the debonding restricted model, where no gap formation between buffer and IPyC layers is permitted, and the debonding enabled model, where the gap between those layers is created. The inputs of the simulated models are based on the irradiation conditions from the AGR-1 experiment. The research included simulations on spherical and aspherical fuel types. Under the specific temperatures and fluences of the AGR-1 irradiation experiment, and the Bison simulations, it was concluded that the most common scenario is a gap formation along the buffer-IPyC interface, while the least possible scenario is the situation where there is no gap formation at the buffer-IPyC junction. The computational results confirmed that the sphericity of the fuel influences the thickness of the gap that occurs at the buffer-IPyC junction, in a way that with increasing aspect ratio the gap thickness increases. The results obtained for spherical and aspherical fuel are nearly identical. Finally, performed simulations match conclusions observed from the AGR-1 experiment, which as such shows that the Bison code is a good computational method for simulating the TRISO fuel. Future simulations will include the validation of performed research and comparison of the results between Bison and PARFUME codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Simulation of the NuScale SMR and Investigation of the Effect of Load-Following on Component Lifetimes

The NuScale SMR has been modelled using the Virtual Environment for Reactor Applications (VERA) multiphysics environment and the results compared with the publicly reported data in the Design Certification Application (DCA). The results show an excellent agreement for the compared axial and radial power distributions, temperature coefficients of reactivity, boron and control rod worths, and fast neutron flux. This NuScale model is then used to investigate the effect of different operational modes on reactor components to determine how flexible load-following operation may affect control rod and reactor pressure vessel (RPV) lifetimes. The control rod degradation is confirmed to primarily affect the silver-indium-cadmium (AIC) rod tip. The degradation rate is observed to follow a non-linear function of core power level where the increase in degradation decreases with insertion depth. For the variation in core power levels expected with current load-following schemes, the total control rod degradation is found to be mild, at 5-10% of usable life per cycle for a reactor operating at <80% power. Nonetheless, this enables load following strategies to be confirmed and/or modified to ensure that control rods do not need to be replaced during the 60+ year life of the reactor. The RPV degradation was found to be almost directly proportional to the core power level and was not overly sensitive to flux shape perturbations. Future work is planned using these damage functions to optimize operation over multiple NuScale SMR units and develop strategies for prognostics and health management.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Investigations on the Thermal-hydraulic Behavior of Accident Tolerant Fuel Cladding Materials

Since the 2011 Fukushima accident, significant research has been devoted to developing accident-tolerant fuel (ATF) cladding materials. These investigations have mostly focused on the ability to resist runaway steam oxidation and retain mechanical strength and structural integrity under thermal shocks. However, much remains unknown about the materials’ thermal-hydraulic behavior, particularly under light-water reactor (LWR) operating conditions. Two phenomena that determine safety margins in both normal and off-normal operating conditions - i.e., the critical heat flux (CHF) and Leidenfrost point (LFP) temperature - have not been investigated thoroughly for ATF materials. Surface properties, such as wettability and roughness, are known to influence pool and flow boiling CHF, as well as the LFP temperature. However, little is known about the surface wettability of ATF materials, particularly at LWR pressure and temperature. Little has been done on flow-boiling CHF and quenching heat transfer, especially for droplet quenching. Nonetheless, it is known that CHF in transient conditions, e.g., an exponentially escalating power transient, can be significantly different from that expected in steady-state operation. Very few studies have investigated transient CHF, either on ATF materials or under LWR pressure and temperature. The thorough understanding of transient CHF under prototypical reactor conditions will benefit not only the deployment of ATF but also the upcoming national efforts to study transient ATF behavior in the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory. In summary, as a part of ATF development efforts, there is an urgent need to understand how and how much these materials may affect two-phase heat transfer phenomena in nuclear reactor conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1987

Small reactor power system for space application

A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

Shirbacheh, M.↗