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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Multiphysics analysis of fuel fragmentation, relocation, and dispersal susceptibility–Part 1: Overview and code coupling strategies

The US nuclear energy industry is investigating strategies to increase the reactor operating cycle to 24 months, resulting in peak rod average burnups exceeding the current limit of 62 GWd/tU. This increase will in turn increase the probability of fuel fragmentation, relocation, and dispersal (FFRD) in the event of a loss-of-coolant accident (LOCA). This effort couples multiple codes to (1) evaluate full-core power histories for high-burnup fuel operated in a Westinghouse 4-loop pressurized water reactor, (2) model a postulated large-break LOCA, and (3) calculate the mass of fuel susceptible to FFRD. This paper, the first of three describing the work, focuses on code coupling strategies and FFRD susceptibility calculations. The other two companion papers focus on code-specific designs and analyses. Three codes were used in this work. VERA was used to calculate steady-state power histories, TRACE was used to model the transient thermal hydraulics, and BISON was used to model steady-state and transient fuel performance and cladding failure. Herein, several fuel pulverization models were used to calculate FFRD susceptibility in failed rods. Depending on the cladding failure/fuel pulverization model combination, the core-wide FFRD susceptibility during the postulated LOCA range from 0 to over 5,000 kg.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full-sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask, in part by increasing the efficiency of internal conduction pathways, and also by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above- and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an above-ground configuration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Aluminum Clad Spent Nuclear Fuel Task 3: Sealed and Vented Systems Episodic Breathing and Gas Generation Modeling Plan

This document presents the test plan for Task 3 of the Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Technical Issues Action Plan, which covers modeling and predictions of the generations of corrosive gases within both sealed and vented, subject to episodic breathing, storage systems. This includes aluminum clad spent nuclear fuel (ASNF) that is currently or may in the future be held in dry storage for an extended period. The overall objectives of the Task 3 modeling plan are to: 1. Comprehensively model the combined effects of episodic breathing in vented storage systems/canisters using INL INTEC CPP-603 facility as an example. Both bulk gas phase radiolysis reactions and aluminum cladding surface gas generation reactions will be included into a 3-dimensional computational fluid dynamics (CFD) model to model and predict long-term evolutions of the temperature and concentrations of corrosive gases within unsealed canisters subjected to episodic breathing. 2. Comprehensively model sealed and inert canister storage system using DOE standard canister as an example. 3. Closely integrate with Task 1 and Task 2 for identification of main gas-generation reactions and kinetics and provide simulation results back to Task 1 and 2 to further refine experimental designs. This test plan lays out the technical activities that will be performed to achieve the test objectives described above. Simulations will be performed using the Idaho National Laboratory super computer Falcon.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Considerations for Improving the Technoeconomic Viability of Heat Pipe Microreactors by Employing a Functional Containment Approach

The economic cost of a microreactor is being investigated by leveraging MARVEL cost information in conjunction with some simplified design and analysis activities to present a hypothetical, yet feasible, option for investigation. This bottom-up cost estimate for a hypothetical commercial microreactor will be used to provide guidance to the industry and research communities regarding where priorities should be set to improve the viability of broad microreactor deployment. To support the design and analysis activities required to present a feasible option, some understanding of maximum accidental radionuclide releases and consequences is needed to provide bounding estimates on containment or confinement systems for radionuclide retention. To estimate fission product barrier and retention performance, a gaussian plume dose and dispersion model is employed, with isotopic inventories being propagated between barriers assuming conservative fractional release rates. Demonstrably conservative meteorological assumptions are also assumed. Standard reference release rates are assumed where safety requirements could be easily and economically employed. Results are provided in units of total effective dose equivalent as a function of distance from the release (i.e., the reactor). Initial results indicate that the hypothetical microreactor will likely need to credit the fuel cladding and at least one other barrier, or combination of barriers, to ensure public health and safety as required during a postulated accident. This approach is referred to as “functional containment,” since no single barrier will be relied upon for satisfying regulatory requirements and fundamental safety functions for radionuclide retention. Finally, a set of barrier options are provided so that future economic analyses can select the most viable design path.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1980

Potentialities of TEC topping: A simplified view of parametric effects

An examination of the benefits of thermionic-energy-conversion (TEC)-topped power plants and methods of increasing conversion efficiency are discussed. Reductions in the cost of TEC modules yield direct decreases in the cost of electricity (COE) from TEC-topped central station power plants. Simplified COE, overall-efficiency charts presented illustrate this trend. Additional capital-cost diminution results from designing more compact furnaces with considerably increased heat transfer rates allowable and desirable for high temperature TEC and heat pipes. Such improvements can evolve of the protection from hot corrosion and slag as well as the thermal expansion compatibilities offered by silicon-carbide clads on TEC-heating surfaces. Greater efficiencies and far fewer modules are possible with high-temperature, high-power-density TEC: This decreases capital and fuel costs much more and substantially increases electric power outputs for fixed fuel inputs. In addition to more electricity, less pollution, and lower costs, TEC topping used directly in coal-combustion products contributes balance-of-payment gains.

Morris, J. F.↗
DOE OSTI2023

Considerations for Managing DOE Standard Canisters within an Over-canister as Part of an Integrated Waste Management System

To better enable informed decision making regarding the back end of the nuclear fuel cycle, the Integrated Waste Management Program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring the research into a comprehensive integrated waste management system (IWMS) that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). An important aspect of the IWMS is DOE-managed SNF. DOE and its predecessor agencies have generated, transported, received, stored, and reprocessed SNF at DOE facilities nationwide, and DOE is responsible for managing the SNF currently in its possession. These fuels come from a wide range of reactor types that employ various cladding materials, fuel materials, and enrichments. To enable interim, road-ready dry storage (RRDS) of the wide variety of SNF types found in the DOE inventory, a standardized canister system (i.e., the DOE Standard Canister) was proposed. This robust, welded canister system is designed to confine radionuclides, prevent criticality by precluding content moderation, and satisfy other requirements as part of a larger storage, transportation, and disposal system. While SNF has yet to be loaded into a DOE Standard Canister, DOE Standard Canister designs were included in past storage facility and disposal facility design licensing endeavors. In a renewed effort to evaluate packaging SNF at Idaho National Laboratory (INL) in a RRDS configuration, researchers are planning the RRDS Packaging Demonstration. This demonstration is supplemented by analytical structural, criticality, and material compatibility evaluations that support management of SNF in DOE Standard Canisters, taking advantage of past analysis work to the extent possible. One of the largest differences between the current Packaging Demonstration and past analytical evaluations is the inclusion of an over-canister containing multiple DOE Standard Canisters. For the Packaging Demonstration, DOE Standard Canisters loaded with SNF are planned to be placed in a larger diameter over-canister. The sealed over-canister could then be placed in a storage overpack for onsite storage, or in a transportation overpack for shipment to an offsite storage location or disposal site once one becomes available. This paper examines the relevant considerations and provides a preliminary evaluation of integrating the over-canister configuration into the storage, transportation, and disposal processes of the overall waste management system. For storage and transportation, the over-canister can be considered analogous to a multi-purpose canister (MPC) for commercial SNF. For disposal, the DOE Standard Canisters could be removed from the over-canister and placed in a co-disposal waste package with canisters containing vitrified HLW similar to configurations examined previously, or the sealed over-canisters might be capable of direct disposal in a waste package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Irradiation performance of nonfertile (Pu-MA-Zr) fast reactor metal fuels

This work was part of a program begun in 2001 to develop advanced nuclear fuels, originally as carriers for plutonium and minor actinides (neptunium, curium, and americium) taken from spent commercial light-water reactors (LWR) so that the plutonium and minor actinides could be ‘burned’ or transmuted in an accelerator or a fast nuclear reactor. A central part of these experiment programs has been the development of advanced fast reactor fuels, because a fast reactor was considered the most efficient vehicle to transmute the actinide waste products, and metallic fuels is a central focus of these tests. An experiment design was developed in which a thermal test reactor, the Advanced Test Reactor (ATR), was used to test small fuel pin prototypes, by creating areas in the core shielded by cadmium filters to produce a largely epithermal and fast neutron spectrum environment in which the pins could be irradiated. The results of non-fertile metallic fuel (no uranium) tests are presented here. Pu-Am-Np-Zr fuels were irradiated to fission densities up to 33 × 10 20 fission/cm 3 and Pu-239 depletions of up to 39%. The depletions were created by roughly 2/3 by fission and 1/3 by transmutation neutron capture. Up to five fuel ‘rodlets’ were irradiated in three sealed capsules stacked axially in the core, and the peak cladding temperatures ranged from 300°C to 500°C, depending on axial location as those near the core centerline are operating hotter and to higher fission densities. Several post-irradiation examinations (precision gamma scanning and fission gas release) were similar to other historical metal fuel experiments in fast reactors. However, optical metallography indicated that two of the rodlets had breached. The exact reasons are unclear. Due to the design of this irradiation experiment a rodlet breach could have increased the temperature in others in the same capsule by contaminating the thermal gap helium with heavier and less conductive fission product gases. Finally, some of those rodlets showed high amounts of fuel/cladding chemical interaction (FCCI).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Conjugate Heat Transfer Modeling of Salt-Filled Fuel Pins for Stable Salt Reactor Safety Analysis

The Stable Salt Reactor (SSR) combines the proven structural design of light water reactor fuel assemblies with the inherent safety and fuel-cycle advantages of molten salt technology. In its fast reactor configuration, the SSR utilizes recycled nuclear waste as fuel, sealed within narrow salt-filled fuel pins and cooled by a surrounding liquid salt coolant. Reliable transfer of heat from the molten fuel salt through the cladding to the external coolant is essential for both reactor safety and performance. This work investigates conjugate heat transfer (CHT) in the SSR’s salt-filled fuel pins using NekRS, a high-fidelity spectral element computational fluid dynamics (CFD) solver. The analyses capture internal natural convection within the molten fuel salt and external forced convection in the coolant, under steady-state and transient operating conditions. Parametric studies evaluate how variations in reactor power and coolant flow rate influence heat transfer distributions and system response. The high-fidelity CFD results are time-averaged and post-processed for direct comparison with moderate-fidelity Reynolds-averaged Navier–Stokes (RANS) models, and for the development of reduced-order models within the SAM system code. These validated models support fast-running safety analyses of normal and off-normal transients, improving predictive capability for key safety margins. By integrating advanced CFD with system-level safety tools, this study strengthens the modeling framework for SSR design, reduces uncertainty in molten salt CHT simulations, and accelerates the engineering and licensing of next-generation nuclear reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Challenges of SLOWPOKE-2 HEU Irradiated Fuel Transportation from Canada to USA - 19123(2)

The SLOWPOKE-2 (Safe Low-Power Kritical Experiment) was designed by AECL (Atomic Energy of Canada Limited) in the late 1960's. The core consists of HEU (highly enriched uranium) fuel with an enrichment of 93%. The fuel cage typically has approximately 300 fuel pins assembled in a triangular pitch between the two plates. The fuel pins are made from uranium-aluminum alloy cladded with aluminum. The SLOWPOKE-2 reactor is rated at a nominal 20 kW. with varying neutron flux, maximum being 1 E12 n/cm{sup 2}/s. The mass of the fuel cage is 5.06 kg. The life of the core can be extended by adding layers of beryllium reflector. The transportation flask for the irradiated fuel is a Type B(U)F. The shipment classification is UN3328 Class 7. To date, 4 SLOWPOKE-2 reactors have been decommissioned. The fuel from all these reactors was transported to the U.S. The CNSC (Canadian Nuclear Safety Commission) joint Canada-United States guide for approval of Type B(U) and fissile material transportation packages (RD-364) satisfies the requirements stipulated in International Atomic energy Agency (IAEA) TS-R-1 1996 edition (revised) and is acceptable to the DOT (Department of Transportation) and NRC (Nuclear Regulatory Commission) for complying with U.S. code of federal regulations 10 CFR Part 71. The export license is issued by the CNSC. The transportation licence is issued by CNSC after approval from TC (Transport Canada) and also by DOT after approval from NRC. The final home of the HEU SLOWPOKE-2 irradiated fuel is in Savannah River Site of DOE (Department of Energy). The transportation licence requires that the Type B(U)F flask is structurally sound in case of a crash. The US DOT provides a CAC (Competent Authority Certificate) which describes the irradiated fuel specifications that can be transported in the designated flask. The U.S. DOT also requires a leak test certification in accordance with ANSI N14.5-2014 to verify a leakage rate not to exceed 1 E-7 std-cm{sup 3}/s. As most SLOWPOKE-2 facilities are not located on the existing approved routes for nuclear fuel transportation, a route survey needs to be performed and submitted to the NRC. The route survey report is assessed by the NRC and issues a route approval for the transportation of the fissile material. This technical paper will discuss the detailed processes required to obtain approvals in order to transport the HEU SLOWPOKE-2 irradiated fuel from Canada to U.S. based upon recent experiences of decommissioning SLOWPOKE-2 reactors in Canada. It will provide details on the requirements of various regulators in Canada and the U.S. that govern and regulate this shipment including transportation, safety and security. It will also present lessons learned from previous experiences that can be utilized for best practice on other similar projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Determination of He and D permeability of neutron-irradiated SiC tubes to examine the potential for release due to micro-cracking

Driven by the need to enlarge the safety margins of light water reactors in both design-basis and beyond-design-basis accident scenarios, the research and development of accident-tolerant fuel (ATF) has become an importance topic in the nuclear engineering and materials community. Continuous SiC fiber-reinforced SiC matrix ceramic composites are under consideration as a replacement for traditional zirconium alloy cladding owing to their high-temperature stability, chemical inertness, and exceptional irradiation resistance. Among the key technical feasibility issues, potential failure of the fission product containment due to probabilistic penetrating cracking has been identified as one of the two most critical feasibility issues, together with the radiolysisassisted hydrothermal corrosion of SiC. The experimental capability to evaluate the hermeticity of SiC-based claddings is an urgent need. In this report, we present the development of a comprehensive permeation testing station established in the Low Activation Materials Development and Analysis laboratory at Oak Ridge National Laboratory. Preliminary results for the hermeticity evaluation of un-irradiated monolithic SiC tubes, uncoated and coated SiC/SiC composite tubes, and neutron-irradiated monolithic SiC tubes at room temperature are exhibited. The results indicate that this new permeation testing station is capable of evaluating the hermeticity of SiC-based tubes by determining the helium and deuterium permeation flux as a function of gas pressure at a high resolution of 8.07 x 10 -12 atm-cc/s for helium and 2.83 x 10 -12 atm-cc/s for deuterium, respectively. The detection limit of this system is sufficient to evaluate the maximum allowable helium leakage rate of lab-scale tubular samples, which is linearly extrapolated from the evaluation standard used for a commercial as-manufactured light water reactor fuel rod at room temperature. The un-irradiated monolithic SiC tube is hermetic, as is manifested by the un-detectable deuterium permeation flux at various feeding gas pressures. A large helium leakage rate was detected for the uncoated SiC/SiC composite tube exposed to atmosphere, indicating it is inherently not hermetic. The hermeticity of coated SiC/SiC composite tubes is strongly dependent on the coating materials and the preparation of the substrate SiC/SiC composite samples. To simulate the practical application environment, monolithic CVD SiC tubes were exposed to neutron irradiation at the High Flux Isotope Reactor under high heat flux from the internal surface to the external surface. Although finite element analysis and resonant ultrasound spectroscopy measurement indicated that the combined neutron irradiation and high heat flux gave rise to a high probability of cracking within the sample, the hermeticity evaluation of the tested sample still exhibited gas tightness, emphasizing that SiC cracking is inherently a statistical phenomenon. The developed permeation testing station is capable of measuring the gas permeation flux in the range of interest with full confidence based on the presented results. It is considered a critical pre- /post-irradiation examination technique to characterize SiC-based cladding materials in asreceived and irradiated states to aid the research and development of ATF.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Design of Laboratory System to Recover Tritium from Zirconium Recycle and Plan to Test System

The objective of the tritium separations task is to develop a viable process to separate tritium from the zirconium recycle off-gas stream. In support of this long-term objective, the goal of the present work is to design a test system and prepare a test plan for an alternate tritium capture method suitable for, and compatible with, recycle of the reactive gas (i.e., chlorine) in the zirconium recycle process. The zirconium recycle process is under development to recover zirconium from the Zircaloy® cladding component of used nuclear fuel. Small quantities of all isotopes of hydrogen (all H) are trapped in the Zircaloy as zirconium hydride. Processing releases and converts this hydrogen to hydrogen chloride, including tritium-substitu ted hydrogen chloride, 3 HCl (a.k.a., TCl). This compound of tritium must be removed from the zirconium recycle process off-gas and sequestered to satisfy regulations governing emissions of tritium.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Report on Exploration of New FeCrAl Heat Variants with Improved Properties

Mechanical and thermophysical properties of wrought FeCrAl base alloys with a concept of accident tolerant fuel (ATF) design for light water reactors (LWR) have been explored. A Gen. II wrought FeCrAl alloy C26M, nominally Fe-12Cr-6Al-2Mo-0.2Si-0.03Y in weight percent, is of interest in the current activities, which was downselected as one of the candidate ATF cladding materials. Major target of the current study is to find the effects of various minor alloying additions and their combinations on potential improvement of thin-wall, seamless tube producibility, without losing the expected mechanical and thermal properties in the final products.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

TEV-3776 Hybrid – ZIRCEX Process Demonstration Siting Study

The purpose of this technical evaluation (TEV) is to detail the siting considerations for supporting pre-conceptual design of the engineering-scale Hybrid – ZIRCEX Process (formerly known as Hybrid Uranium Recovery Demonstration [HURD]) demonstration. This process represents the first-of a kind integration of a select group of technologies meant to recover high enriched uranium (HEU) from zirconium and aluminum clad used nuclear fuel (UNF) for down-blending to produce high assay low enriched uranium (HALEU).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Hydrogen Permeation in FeCrAl APMT Alloy for Accident Tolerant Fuel Cladding

Iron-chromium-aluminum (FeCrAl) alloys such as APMT (advanced powder metallurgy tubing) are candidate materials to replace zirconium alloys for the light water reactor fuel cladding. This alloy meets the requirements to be a material more tolerant of high-temperature accidents than the current zirconium alloys. One concern is that the use of FeCrAl may result in an increase in tritium presence in the coolant compared to the current design. The aim of the current research was to obtain effective diffusion coefficients (D eff ) for hydrogen through APMT using the Devanathan-Stachurski cell. Results showed that at 30°C the D eff value was 2.8 × 10 -8 cm 2 /s. Results also showed a linear relationship between the permeated hydrogen flux and the inverse of the test specimen thickness, which demonstrated the validity of the permeation measurements.

36 MATERIALS SCIENCE↗
NASA NTRS1987

Curved channel MCP improvement program

Blowholes and blemishes were determined to start at two stages of manufacturing. Sperical blowholes resulted from trapped gas between the high melting temperature bond glass and the MCP wafer. During thermal processing, the trapped gas expanded and displaced the softened channel glass to form a spherical inclusion. This defect was eliminated by grinding the prefritted bond wafer and channel plate wafer to a flatness which ensured intimate contact prior to fusion. Elliptical blowholes or blemishes were introduced during the fiber draw stage. Contaminants trapped between the core bar and clad tubing volatized providing large quantities of expanding gas. These pockets of gas became elongated to an ellipsoidal shape during fiber draw. Special cleanliness procedures were developed for the grinding, polishing, and acid etching of core bars. Improvements in channel curvature fabrication were implemented. The design of the shearing fixture was evaluated. A new design was developed which eliminated an off-axis moment. The shearing furnace design was evaluated. Steady state thermal conditions instead of thermal transient conditions were determined to reduce curvature nonuniformity.

Laprade, Bruce N.↗
DOE OSTI2016

Monolithic U-10Mo Fuel Development for U.S. High Performance Research Reactors

To support the U.S Department of Energy (DOE), National Nuclear Security Administration (NNSA) mission to reduce the threat posed by the civilian use of highly enriched uranium (HEU) worldwide, the US High Performance Research Reactor (US HPRR) fuel development program is pursuing development and qualification of a new high-density monolithic U-Mo fuel to facilitate conversion of five higher power research reactors located in the U.S. (ATR, HFIR, NBSR, MIT and MURR) and one critical facility (ATR-C) from HEU to low enriched uranium (LEU). These reactors require higher uranium fuel density than is achievable with dispersion fuel systems used to convert lower power research reactors. The down selected fuel system consists of U-10Mo alloy fuel foils having a thin Zr diffusion barrier interlayer, clad in 6061 Al alloy by hot isostatic pressing. In order to support fabrication development, fuel performance evaluations, generic fuel qualification through the Nuclear Regulatory Commission (NRC), and eventual reactor licensing, a series of irradiation experiments and testing campaigns are being planned. The tests are designed to provide data needed for fabrication process down selection, to establish acceptable fuel performance under normal operating conditions and anticipated transients, and to demonstrate successful scale-up to prototypic reactor-specific fuel element designs. A comprehensive testing program is being executed as part of the US HPRR fuel development program to develop and qualify a new monolithic UMo fuel system to allow conversion of five high power research reactors and one critical facility within the US in support of the NNSA mission to convert research reactors worldwide from HEU to LEU fuels. A significant outcome of this program will be the qualification and eventual licensing of a new plate-type fuel system for use in research reactors through the NRC for the first time in nearly 3 decades. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

A Revised Capsule Design for the Accelerated Testing of Advanced Reactor Fuels

The Advanced Test Reactor (ATR) has been used successfully for the testing of fast reactor fuel for nearly two decades. These successes have been in spite of numerous challenges for testing fast reactor fuel in the ATR (a thermal spectrum reactor), but the solutions to those challenges have resulted in excessively long irradiation times (~10 years) for high-burnup targets as well as experiments that are highly sensitive to fabrication tolerances and eccentricities. This paper presents a solution to the problems of extended irradiation times and fabrication sensitivities. Thermal and neutronic analyses were performed to show that a reduced-diameter fuel pin with an equivalent linear heat generation rate can provide a prototypic thermal profile (peak centerline and inner clad temperature) along with a near-prototypic power profile within the ATR thermal spectrum. This allows the experiment to reach a high burnup in an expeditious timeframe compared to traditional ATR fast fuel irradiations. In addition, problems with fabrication sensitivities were addressed by introducing a double-encapsulated experiment that pushes the high heat flux helium gap farther away from the fuel pin. Fuel pin position eccentricities are also mitigated by using a large sodium bond between the pin and capsule fuel. The advantages and potential pitfalls of this revised design are discussed, including the effect of length scales on fuel system behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

A Eulerian three-fluid flow framework for the study of fuel dispersal behavior under loss-of-coolant accident conditions

As the nuclear industry explores new fuel designs to accommodate increased burnup, studying fuel behavior during loss-of-coolant accidents is essential for ensuring the safe operation of light-water reactors. Here, this study presents a computational model to simulate the intricate three-phase flow dynamics of fuel dispersal following cladding breach, where solid fuel particles and carrier fission gases interact with the surrounding liquid or gas medium. The model utilizes a Eulerian three-fluid framework to depict the phase interaction within interpenetrating continua, treating fragmented fuel as granular material. Closure relationships for bulk-flow properties of the solid phase and detailed interfacial exchange terms in high-temperature, high-speed gas-solid-liquid flow scenarios are proposed. The model was validated using adiabatic experimental results in the literature, focusing on characterizing fuel dispersal behavior during loss-of-coolant accidents conditions. Settlement characteristics of the solids, quantified by the probability distribution of equivalent particles, closely matched probability density functions reported in experimental studies. Key highlights of this study include the theoretical description of a coupled three-phase flow with phase change, interpretation of the high-pressure boundary conditions and insights into the transient behavior of gas-solid-liquid phase dynamics during a simulated high-speed dispersal event.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗