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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Modeling Approach for the Aluminum-clad Dry Storage Pilot using HFIR Fuel

To confirm that the dry storage of aluminum-clad research reactor spent nuclear fuel (ASNF) will remain within the safety envelope after applied drying schemes and that the resulting evolution of the gas space composition, temperature, and pressure conditions are understood, a dry storage pilot project is being established. The pilot will incorporate an instrumented lid for discrete interval or for on-demand gas composition and temperature monitoring of two DOE Standard Canisters (DSCs) loaded with three High Flux Isotope Reactor (HFIR) inner cores per DSC. Each DSC would be subjected to a separate alternative candidate drying scheme. Canisters will undergo 1 to 5 years of monitoring, including internal temperature and gas sampling to track pressure and composition changes. This report outlines the approach for modeling the ASNF-in-canister behavior in terms of evolving gas space conditions for the ASNF dry storage pilot using HFIR fuel. The ASNF has an adherent surface oxyhydroxide layer comprised of boehmite/bayerite that generates hydrogen when subjected to irradiation. Three-dimensional multi-physics computational fluid dynamics simulations will be executed to compute the thermal field within the DSC and provide inputs to a chemical model employed to compute pressure buildup as hydrogen is generated in the system. Implemented in Cantera, the chemical model solves gas phase and aluminum oxyhydroxide surface-mediated radiolysis reactions. Gas phase reactions are sourced from Wittman and Hanson (2015), whereas surface-mediated reactions are incorporated by fitting experimental data using an optimization algorithm (Abboud, 2023). Water radiolysis reactions from Wren and Ball (2001) are adopted with modifications as described in Abboud (2023c). Understanding the effect of the hydrogen buildup over time is important for long-term storage safety considerations. Modeling results will include the canister pressure, temperature, and composition evolution from the initial helium backfill with the addition of radiolytically-evolved chemical species (e.g., hydrogen and oxygen). The specific HFIR cores for the pilot program have not yet been selected, and the overall design is still in development. The CFD-chemical model used for this work will be based on prior models with necessary updates to allow for improved accuracy and efficiency. The experimental data obtained from the HFIR demonstration will be used to improve and validate the computational models to predict the ASNF-in-canister behavior.

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

R&D with NREL's Davison Circulating Riser System

The Davison circulating riser (DCR) is the National Renewable Energy Laboratory's (NREL's) pilot-scale recirculating riser reactor system. Coupled with an upstream fluid-bed pyrolysis system, it permits evaluation of catalytic cracking of biogenic feedstocks, as well as coprocessing of biogenic and fossil feedstocks in a commercially relevant pilot fluid catalytic cracker.

09 BIOMASS FUELS↗
DOE OSTI2017

BWR Full Fuel Assembly Testing and Validation

This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CLS.P15.10: BWR Full Fuel Assembly Testing and Validation. The extension of existing boiling closure designed for Pressurized Water Reactors (PWRs) to the high void fraction condition of Boiling Water Reactors (BWRs) is one of the challenges for Multiphase Computational Fluid Dynamics (M-CFD) applications. In the efforts aimed at developing the advanced boiling closure for BWRs, the baseline closure model, so-called ‘Zero’ closure, has been developed based on the GEN-I closure for PWRs which was used as a starting point. Since this simplest closure model is robust but does not contain any explicit description of the complex high void fraction mechanisms, the applicability of M-CFD needs to be evaluated before developing advanced closures for BWRs. To support this need, the high fidelity M-CFD simulation with Zero closure has been conducted for the BWR full fuel assembly. The capabilities of M-CFD with Zero closure were evaluated against the international OECD/NRC BFBT benchmark data. The scalability of the M-CFD application to the nuclear fuel assembly was demonstrated with the INL High Performance Computing (HPC) system. This report presents the recent view of the best practices on M-CFD for BWRs to provide a useful reference to identify and overcome common challenges in the M-CFD applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Coupled Calculations of the NuScale FIV Tests – Pre-test Simulations and Sensitivity Analysis

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict FIV based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the conceptual design stage. Over the past several years, the Reactor Product Line has developed a FIV capability in SHARP. SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. In previous reports we have discussed the implementation and validation of an FIV capability for helical steam generators and fuel assemblies. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo for the NuScale FIV tests. In the turbulent fretting regime one-way coupling is judged to be sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators per the low flows expected in integral PWRs. The simulations precede the tests, as experimental data was not available at the delivery of this report. Several assumptions had to be made on the support-tube interaction and multiple tests and sensitivity analysis have been performed. Additional work remains to be done and it is discussed in the future work section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1991

Optimum rocket propulsion for energy-limited transfer

In order to effect large-scale return of extraterrestrial resources to Earth orbit, it is desirable to optimize the propulsion system to maximize the mass of payload returned per unit energy expended. This optimization problem is different from the conventional rocket propulsion optimization. A rocket propulsion system consists of an energy source plus reaction mass. In a conventional chemical rocket, the energy source and the reaction mass are the same. For the transportation system required, however, the best system performance is achieved if the reaction mass used is from a locally available source. In general, the energy source and the reaction mass will be separate. One such rocket system is the nuclear thermal rocket, in which the energy source is a reactor and the reaction mass a fluid which is heated by the reactor and exhausted. Another energy-limited rocket system is the hydrogen/oxygen rocket where H2/O2 fuel is produced by electrolysis of water using a solar array or a nuclear reactor. The problem is to choose the optimum specific impulse (or equivalently exhaust velocity) to minimize the amount of energy required to produce a given mission delta-v in the payload. The somewhat surprising result is that the optimum specific impulse is not the maximum possible value, but is proportional to the mission delta-v. In general terms, at the beginning of the mission it is optimum to use a very low specific impulse and expend a lot of reaction mass, since this is the most energy efficient way to transfer momentum. However, as the mission progresses, it becomes important to minimize the amount of reaction mass expelled, since energy is wasted moving the reaction mass. Thus, the optimum specific impulse will increase with the mission delta-v. Optimum I(sub sp) is derived for maximum payload return per energy expended for both the case of fixed and variable I(sub sp) engines. Sample missions analyzed include return of water payloads from the moons of Mars and of Saturn.

Zuppero, Anthony↗
DOE OSTI2016

Overview of the Salt at Wall Thermal Exchanges (SWATH) Experiment

Molten Salt Reactors have attracted increased attention in recent years because of the design and safety possibilities offered by the use of a liquid fuel. The European project SAMOFAR1 (2015-2019) is currently coordinating the research efforts on the Molten Fast Salt Reactor (MSFR) concept in Europe. The MSFR is a fast-spectrum breeder reactor with a large negative power coefficient that can be operated in a Thorium fuel cycle. The fast spectrum allows the reduction of the reprocessing requirements and a better reactor breeding ratio. Other advantages of the MSFR design are the possibility for actinide burning and extending fuel resources, on-line fuel loading and reprocessing and the use of novel passive safety systems such as the fuel salt draining system. The reference MSFR concept is a 3000 MW(th) reactor with three different circuits: the fuel circuit, the intermediate circuit and the power conversion system. The main components of the fuel circuit are the fuel salt which serves as fuel and coolant, the core cavity, the inlet and outlet pipes, the gas injection system (not shown in the figure), the salt-bubble separators, the fuel heat exchangers and the pumps. A lithium fluoride salt has been selected as the fuel matrix. The initial composition (non-irradiated) of the MSFR fuel salt is a mixture of a lithium fluoride, thorium fluoride salts and actinides fluoride (LiF-ThF{sub 4}-{sup 233}UF{sub 4} or LiF-ThF{sub 4}- {sup enr}UF{sub 4}-(Pu-MA)F{sub 3}), with the proportion of LiF fixed at about Development of accurate numerical models for the MSFR can only be accomplished by the use of a multiphysics approach which allows taking into account the coupling between relevant neutronics, thermal hydraulics and thermomechanics phenomena. As it is discussed in the next section, such a reactor model requires further progress on the knowledge of molten salt thermal hydraulics, in particular regarding the accuracy of the fuel salt Computational Fluid Dynamics (CFD) models. The Salt at Wall: Thermal Exchanges (SWATH) experiment is being designed to contribute to this particular point. SWATH is one of the experimental activities of the European H2020 SAMOFAR project and includes both experimental and numerical developments. The main objectives of SWATH are to: (i) Improve molten salt numerical models used for design and safety studies and (ii) Demonstrate the working principles of the cold plug device. The cold plug device is a key safety component of the salt draining system of the MSFR. The cold plug is intended to passively melt during accidental conditions thus allowing the flow of the nuclear fuel from the core cavity to the fuel salt draining tanks by gravity forces. The fuel salt draining tanks are designed to ensure that adequate margins are obtained regarding the fuel cooling capabilities and the reactivity control. The next sections provide an overview of the current challenges in molten salt thermal hydraulics modeling and the SWATH experiment design. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Microreactor Agile Nonnuclear Experimental Testbed Test Plan

Microreactors are an attractive technology option for kick-starting nuclear innovation if they can be operated at high temperature, yielding high power conversion thermal efficiencies comparable or better than in commercial light water reactors. Microreactors are currently the smallest variation of Small Modular Reactors (SMRs). SMRs are “newer generation reactors designed to generate electric power up to 300 MWe and whose components and systems can be shop-fabricated and then transported as modules to the sites for installation as demand arises.” (IAEA, 2016). Vendors are developing microreactor designs to provide an affordable, potentially mobile source of electricity - see Fig. 1 for an example of a microreactor on a semi-truck. Various microreactor designs are possible including heat pipe- and gascooled options, which are the focus of the nonnuclear testing described in this document. In heat pipe microreactors, high-temperature heat pipes using liquid sodium or potassium working fluid transport fission heat from the core to a heat removal section which in turn transfers heat to the power conversion system working fluid. In a gas-cooled design, He or other gas will flow through a solid monolith of material and transfer heat as the temperature of the gas increases through a heat exchanger to a power conversion unit. The logistics of all these processes will be examined and tested through a series of articles at the nonnuclear test bed at Idaho National Laboratory (INL), the Microreactor Agile Nonnuclear Experiment Testbed (MAGNET) facility. Microreactors designed to produce power of 0.1-20 MWt offer the potential for more affordable nuclear energy for a range of applications. In a heat pipe microreactor, heat pipes, fuel rods, and/or moderator are intermixed in the reactor core assembly. Heat pipes extend from the core region into the heat removal section where the power conversion unit working fluid flows through holes or channels, transferring heat from the heat pipes to the working fluid. In a gas-cooled microreactor, gas flows through the solid monolith region and up into the heat exchanger region, transferring heat to the working fluid. For initial testing, the heat removal working fluid can be a low pressure gas for testing that addresses thermal stresses. In the final application, heat addition to the power conversion working fluid typically occurs at high pressure, supporting operation of an air-Brayton, supercritical CO 2 (SCO2), or He-recuperated Brayton cycle. Various stages of the steps above will be demonstrated through the tests described in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2017

Investigation of a Tricarbide Grooved Ring Fuel Element for a Nuclear Thermal Rocket

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the temperature limitations of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment.

Taylor, Brian D.↗
NASA NTRS2018

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nu- clear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and de ne a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Taylor, Brian↗
NASA NTRS2018

Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are discussed. The authors demonstrated success in reaching desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and define a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

Taylor, Brian↗
DOE OSTI2020

Using Resin to Reduce Silica in Borated Fluid - 20072

The Electric Power Research Institute (EPRI) strongly recommends the concentration of silica in the reactor coolant be limited, potential ingress be carefully controlled, and frequently monitored to assess potential silica deposits on fuel cladding surfaces. Boraflex is a product with silicone rubber encasing a neutron absorber for criticality control, and it was installed in the majority of nuclear power plant spent fuel pools. Over time and in the presence of radiation, the silicone rubber lining degrades into silica (SiO{sub 2}) and hydrogen gas. Silica is somewhat soluble in water and spent fuel pool water communicates with other plant systems during refueling outages, leading to a buildup of silica in the reactor coolant and connected systems. Due to the prevalence of Boraflex racks in spent fuel pools, the presence of silica is a chronic nuclear industry concern. To address this concern, many nuclear plants have installed reverse osmosis skid systems to remove the silica from their spent fuel pools and, by extension, from reactor coolant. These skids are expensive to install and maintain. Arcadis personnel demonstrated that an iron-impregnated resin could be used to remove silica from borated systems without impacting the water quality. The resin was readily available, inexpensive, and easy to install in existing plant equipment. During a nuclear power plant's refueling outage the iron-impregnated resin proved to be effective in reducing the need for feed and bleed operations, with concurrent cost savings and reduction in required outage water management activities. Thus, the following savings were realized: - Eliminated the need to purchase, install, and maintain a reverse osmosis system avoiding more than $10 million in up-front capital cost and ongoing operating and maintenance costs. - Removed 9.57 kilograms (21 pounds) of silica dioxide from the primary system fluid. - Reduced the volume of feed and bleed from 70,000 gallons (2.65 E+05 liters) to 35,000 gallons (1.32 E+05 liters). - Saved $380,000 in water production and disposal costs. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2015

Apparatus and method for solid fuel chemical looping combustion

The disclosure provides an apparatus and method utilizing fuel reactor comprised of a fuel section, an oxygen carrier section, and a porous divider separating the fuel section and the oxygen carrier section. The porous divider allows fluid communication between the fuel section and the oxygen carrier section while preventing the migration of solids of a particular size. Maintaining particle segregation between the oxygen carrier section and the fuel section during solid fuel gasification and combustion processes allows gases generated in either section to participate in necessary reactions while greatly mitigating issues associated with mixture of the oxygen carrier with char or ash products. The apparatus and method may be utilized with an oxygen uncoupling oxygen carrier such as CuO, Mn.sub.3O.sub.4, or Co.sub.3O.sub.4, or utilized with a CO/H.sub.2 reducing oxygen carrier such as Fe.sub.2O.sub.3.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Involute Working Group – Validation of CFD Turbulence Models for Steady-State Safety Analysis

There are three research reactors in the world having fuel plates curved as circle-involute (a spiral generated around a circle): The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) located in Tennessee, U.S.A. [ORNL, 2018]; The Laue-Langevin Institute (ILL) High Flux Reactor (RHF) located in Grenoble, France [ILL, 2018]; and, The Technical University of Munich (TUM) Research Neutron Source Heinz Maier-Leibnitz (FRM II) located in Garching, Germany [TUM, 2018]. All three reactors are currently using Highly Enriched Uranium (HEU, 235U/U ≥ 20 wt. %) as fuel and all three are actively engaged in activities to convert to Low Enriched Uranium (LEU, 235U/U < 20 wt. %) fuel. More information on these reactors can be found in Section two. For various reasons, these reactors have expressed interest in using Computational Fluid Dynamics (CFD) tools to perform their Steady-State Thermal-Hydraulic (SSTH) safety calculations. Using CFD tools for SSTH analysis represents generally a significant departure from traditional methods and requires thorough verification and validation (V&V) to be accepted by regulators. This is why Argonne National Laboratory (ANL) and the involute reactors formed an informal group to help each other in this endeavor. This so-called Involute Working Group (IWG) aims at supporting the qualification of CFD tools for SSTH safety calculations. Activities include benchmarking, code-to-code comparison, Verification and Validation (V&V) as well as technical support. More information on the IWG and the reasons why involute-plate reactors are interested in CFD can be found in Section three. The present study aims at testing three turbulence models commonly implemented in commercial software (namely K-Epsilon, K-Omega and Reynolds-Stress) and assessing their relevance for involute-plate reactor application. An experiment performed by Gambill and Bundy in 1961 was performed to measure heat transfer coefficients in conditions prototypic of the HFIR reactor [Gambill, 1961]. More details on this experiment are provided in Section four. Based on the information available, CFD models of this experiment have been created and the experimental results have been compared to the ones obtained with these CFD models. CFD models and the results of this comparison are provided in Section five.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1993

Analysis of plume backflow around a nozzle lip in a nuclear rocket

The structure of the flow around a nuclear thermal rocket nozzle lip has been investigated using the direct simulation Monte Carlo method. Special attention has been paid to the behavior of a small amount of harmful particles that may be present in the rocket exhaust gas. The harmful fission product particles are modeled by four inert gases whose molecular weights are in a range of 4 131. Atomic hydrogen, which exists in the flow due to the extremely high nuclear fuel temperature in the reactor, is also included. It is shown that the plume backflow is primarily determined by the thin subsonic fluid layer adjacent to the surface of the nozzle lip, and that the inflow boundary in the plume region has negligible effect on the backflow. It is also shown that a relatively large amount of the lighter species is scattered into the backflow region while the amount of the heavier species becomes negligible in this region due to extreme separation between the species. Results indicate that the backscattered molecules are very energetic and are fast-moving along the surface in the backflow region near the nozzle lip.

Chung, Chan H.↗
DOE OSTI2019

Verification and Validation of COMSOL for Heat Transfer in Thin Rectangular Channels Using NACA Test Results

Currently efforts are underway at Oak Ridge National Lab (ORNL) to convert the High Flux Isotope Reactor (HFIR) from high enriched uranium to low enriched uranium fuel. In order to provide confidence to the thermal hydraulics calculations, computational fluid dynamics (CFD) simulations – simulated using COMOSOL – have been employed to calculate the pressure drop and heat transfer in the coolant. While CFD simulations provide well resolved results, turbulence models include a large amount of assumptions and must be validated for use in any safety-based analysis. In HFIR, long thin involute fuel plates are separated by coolant channels of equal size. The coolant channels are best described as thin rectangular channels. Unfortunately, there are few studies verifying and validating CFD for use in thin rectangular channels. This study seeks to bridge that gap by verifying, validating, and comparing various CFD models used in thin channels. To verify and validate CFD for use in thin channels, a series of simple analytic test cases are employed, and MMS is used to further verify the COMSOL tool and mesh geometry. For the validation, a series of tests conducted at NASA were simulated in COMSOL. The results of this study are three-fold. First, it verifies the ability of COMSOL to model heat transfer in thin channels. Second, it addresses the various aspects of CFD modeling that are of greatest concern for heat transfer simulations in the geometry. Third, and finally, it discusses the ability of various CFD models used in COMSOL to accurately predict heat transfer in narrow rectangular channels.

Howard, Trevor↗
DOE OSTI2022

Two-stage combustor

A two-stage combustor having as constituent parts: a partial oxidation reactor, which catalytically converts a hydrocarbon fuel and a first supply of oxidant into a gaseous partial oxidation product; and a deep oxidation reactor having a premixer plenum fluidly connected to a porous heat spreader, which converts the gaseous partial oxidation product to deep oxidation products. In one embodiment, the premixer plenum provides an empty space wherein combustion occurs in flame mode. In a second embodiment, the premixer plenum contains a high pore density foam matrix, absent catalyst, which facilitates holding a flameless combustion downstream within the porous heat spreader. In both embodiments heat produced during combustion is transmitted from the heat spreader to an associated heat acceptor, such as a heater head of a Stirling engine.

Roychoudhury, Subir↗
NASA NTRS1977

Gaseous-fuel nuclear reactor research for multimegawatt power in space

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

Thom, K.↗