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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,531 records · Page 85
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) or plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used in this paper to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. Finally, in comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Space and Defense Power Systems Maintenance Plan

Los Alamos National Laboratory (LANL) has more than forty years of experience in developing, characterizing, testing, and manufacturing plutonium-238 ( 238 Pu) heat sources. The LANL heat source program has made significant contributions to the United States space, defense, and national security missions, as well as the nuclear-powered artificial heart program. Currently, the Office of Space and Defense Power Systems (SDPS), located in the Nuclear Energy Group (NE-75), funds LANL for the infrastructure and processes used to store, recover, manufacture, and test 238 PuO 2 heat sources. LANL is an approved supplier of the General Purpose Heat Source (GPHS), which is a 238 PuO 2 Fueled Clad designed to provide 62.5 thermal watts. The Light Weight Radioisotope Heater Unit (LWRHU) is another 238 PuO 2 fueled heat source manufactured at LANL. The LWRHU is designed to provide one thermal watt in various locations on a spacecraft. LANL also supports the manufacture of other heat sources for various customers.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Verification of the Coupled Neutronics/Thermal-Hydraulics Code TANSY against the MSFR Benchmark

The MSRs is a liquid-fuel reactor in which uranium, fission products and actinides are dissolved in the molten salt. The molten salt reactors are featured by circulating fuel, low pressure and high temperature. The MSRs meets many of the future goals of nuclear technology, in particular for what concerns as improved sustainability and unique characteristics in terms of actinide burning. The fluid nature of fuel in MSRs gives extra flexibility in reactor design, fuel fabrication and recycling. But because of the application of fluid fuel, MSRs have several special characteristics compared with the traditional solid-fuel reactors. Firstly, the transport of the delayed neutron precursors by the fluid flow reduces the contribution of delayed neutrons to the chain reaction.. Secondly, the fact that the fuel is dissolved in the coolant rather than separated from the coolant by the claddings results in a much stronger coupling phenomenon between the neutronics and thermal-hydraulics. Thirdly, the shape of the fuel is determined by the container because of the characteristics of fluid, the unstructured meshes are required to model the complex geometry accurately. However most of the conventional reactor codes were developed for solid-fuel reactor and therefore not capable of taking liquid fuel flow into account. To address all these particular modeling issues of MSRs, a multi-physics code named TANSY has been developed in our laboratory. In this paper, the benchmark problem based on the MSFR design was calculated to verify the code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Neutron Absorber Considerations for the DOE Standardized Canister

The Department of Energy (DOE) standardized canister is being re-evaluated to support the management of non-commercial spent nuclear fuel. The standardized canister would allow a broad range of DOE-managed fuels to be loaded, stored, transported, and disposed of while avoiding the need to re-open the canister after initial loading, thus preventing repackaging of the material. A key safety function of this package is to ensure the material maintains a sub-critical configuration for storage, transportation, and disposal. DOE-managed fuels come from a wide range of reactor types, with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties resulting in a large diversity of reactor and fuel designs. To adequately address the diversity of DOE spent nuclear fuel (SNF), a number of basket were designed. The basket material needed to have the ability to ensure subcriticality, be structurally sound, and be corrosion resistant. One such option that was investigated in the late-1990s through mid-2000s was a corrosion-resistant, nickel-chromium-molybdenum alloy containing gadolinium, which was termed the advanced neutron absorber (ANA). A number of analyses were performed to verify the structural, corrosive properties, and neutron absorption performances of this material as well as to compare other potential materials to be used for the basket. Gadolinium-containing carbon steel and borated stainless steel were also considered and evaluated. Boral was briefly considered, but due to its corrosive properties. Each of the evaluated materials have strengths and weaknesses. ANA is the best thermal neutron absorber but lacks confirmatory research for corrosion and weldability. Gadolinium containing carbon steel provides excellent moderator exclusion capabilities but does not contain a sufficient supply of phosphate to ensure the gadolinium precipitates. To minimize the probability of a critical configuration, it is highly desirable for the gadolinium to precipate the probability that the materials will remain within the waste packages over the postclosure time periods of interest. Borated stainless steel has structural integrity and impressive corrosion resistance in most environments, but can undergo galvanic corrosion when in close proximity/contact with aluminum and may experience less than desirable corrosion resistance. This report provides an extensive list and comparison of the materials evaluated to ensure sub-criticality of DOE standardized canister in all credible scenarios. More analyses need to be performed using new and updated corrosion information to re-evaluate the neutron absorber that best meets the objectives of the DOE standardized canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Progress in Fast Modular Reactor Conceptual Design

The Fast Modular Reactor (FMR) is a 100-MW(thermal) gas-cooled fast reactor being developed by General Atomics Electromagnetic System with the goal of developing a FMR for flexible and dispatchable power to the U.S. electricity market in the mid-2030s. The conceptual design aims to develop and verify simplified design features. These include an inert helium gas coolant, pellet-loaded fuel rods, installations with air cooling as ultimate heat sink, and small and passive heat removal systems. The goal is to ensure the development of a safe, maintainable, cost-effective, and distributed nuclear energy-generating station. The baseline technologies selected to achieve this goal are a helium coolant that is an inert gas with no chemical reaction with structural components, not activated, single phase, enabling high-temperature operation and a high thermal efficiency Brayton cycle; conventional uranium dioxide (UO 2 ) fuel, which is the most widely used and well-known fuel material, capable of high burnup (100 MWd/kg) and a long fuel life; and silicon carbide composite (SiGA®) cladding and internal structures that are chemically inert in the helium environment, exceptionally radiation tolerant, and being derisked by accident tolerant fuel technology development. Further, the reactor was specifically designed with passive safety features, including high-temperature in-core materials and a reactor vessel cooling system consisting of cooling panels of naturally circulating water. The passive safety of the core was confirmed for the depressurized loss-of–forced cooling accident, which showed the peak cladding temperature at ~1600°C during the transient, which is below the current design limit of 1800°C. The conceptual design of the FMR has been conducted for the reactor system, vessel system, generator and turbomachine, instrumentation and control, residual heat removal system, plant service system, and containment, as well as pre-application licensing documents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1974

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

Bowles, K. J.↗
NASA NTRS1997

Engineering design aspects of the heat-pipe power system

The Heat-pipe Power System (HPS) is a near-term, low-cost space power system designed at Los Alamos that can provide up to 1,000 kWt for many space nuclear applications. The design of the reactor is simple, modular, and adaptable. The basic design allows for the use of a variety of power conversion systems and reactor materials (including the fuel, clad, and heat pipes). This paper describes a project that was undertaken to develop a database supporting many engineering aspects of the HPS design. The specific tasks discussed in this paper are: the development of an HPS materials database, the creation of finite element models that will allow a wide variety of investigations, and the verification of past calculations.

SPACECRAFT POWER SUPPLIES↗
DOE OSTI2016

Thermal Modeling of a Loaded Magnastor Storage System at Catawba Nuclear Station

This report represents fulfillment of the M2 Milestone M2FT-16PN080203031, “High Heat Load Thermal Analyses,” under Work Package FT-16PN08020303. The report documents the analysis of the NAC International, Inc. (NAC) Modular, Advanced Generation, Nuclear All-purpose STORage (MAGNASTOR) module at Duke’s Catawba Nuclear Station. As part of the Used Fuel Disposition Campaign of the U.S. Department of Energy, a high-burnup fuel storage demonstration is planned for a storage module in the North Anna Nuclear Station’s Independent Spent Fuel Storage Installation (ISFSI). The storage module selected for this demonstration is an AREVA TN-32B High Burnup cask. The main goals of this proposed test are to provide confirmatory data for model validation and potential improvement, support license renewals and new licenses for ISFSIs, and support transportation licensing for high burnup spent nuclear fuel (EPRI 2014). The focus of the Demonstration test is the performance of the high-burnup fuel. To support the High Burnup Spent Fuel Data Project, Pacific Northwest National Laboratory (PNNL) staff performed a detailed thermal analysis (Fort et al. 2016) of the TN-32B cask. It was found that when conservatisms were removed from the thermal models and especially from the calculated decay heats, the predicted peak cladding temperatures (PCTs) were significantly below the 400°C (752°F) regulatory guidance limit, as specified in Interim Staff Guidance 11, Revision 3, (U.S. NRC 2003).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Nuclear fuel alloys or mixtures and method of making thereof

Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Impact of Molten Gallium on the Microstructure and Corrosion Behavior of Aluminum and Uranium-Aluminum Alloys for Used Nuclear Fuel Reprocessing

Test reactors around the world utilize highly enriched uranium fuel to achieve high neutron fluxes for materials testing. Once spent, the remaining uranium is a valuable resource for subsequent fuel fabrication. However, some of these test reactor cores consist of curved plate-type fuel elements, fabricated using aluminum alloy 6061 (AA6061) cladding to encapsulate a uranium-aluminum alloy (UAlx) fuel matrix. These assemblies require non-standard reprocessing approaches for uranium recovery, as aluminum dissolves readily in acidic solutions, generating large volumes of waste and complicating downstream chemical separations. In this work, we investigate a novel chemical decladding strategy based on the interaction between AA6061/UAlx and molten gallium (Ga). Ga is known to induce severe degradation of aluminum metal through liquid metal embrittlement (LME), even at relatively low Ga concentrations. By penetrating the aluminum crystal lattice, Ga disrupts grain cohesion and facilitates fracture or dissolution of the aluminum matrix. Thermodynamic analysis of the Al–Ga binary phase diagram suggests that Ga may offer a viable pathway to selectively weaken or dissolve the AA6061 cladding, and potentially the aluminum component of the UAlx fuel matrix within. To this end, parametric experiments were performed at 50 °C and 100 °C across a range of Al–Ga atomic fractions. At lower Al fractions, the AA6061 was completely molten after 2 hours of exposure to the Ga metal. In contrast, samples with higher Al fractions (0.9 Al, 0.1 Ga) contained residual solids after 2 hours, which were characterized by microstructural examination using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). These Al-Ga compositions were also evaluated using FactSage thermodynamic modeling to further elucidate the relationship between phase diagram behavior and LME.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Axial Relocation Model Extension in Bison

Fuel rods subjected to a Loss of Coolant Accident (LOCA) in a nuclear reactor can undergo a complex process known as fuel fragmentation, relocation, and dispersal (FFRD). The loss of coolant results in a severe degradation of the thermal heat transfer coefficient on the waterside surface of the cladding. This, coupled with the loss of coolant pressure causes the cladding to distend (balloon) outward. If the cladding distention is large enough, fuel fragmented during irradiation, may relocate from upper regions of the rod into the ballooned region. This mass relocation results in increased localized heating due to the presence of more fuel within the balloon resulting in further clad distention leading to potential rupture of the cladding. In the event that cladding rupture occurs the relocated fragmented fuel may disperse outside of the rupture opening if the size of the fuel fragments is small enough. Fuel performance modeling can be used to help understand the behavior of fuel rods during the LOCA transient. Existing models are particularly limited in regards to FFRD. This report briefly describes the existing state of the art for fuel performance modeling of fragmentation and fuel axial relocation. A more in depth description the axial relocation model implemented into the Bison fuel performance code in FY17 is provided. In this report, the extension of this existing model from a layered 1D framework to a 2D framework that allows for the introduction of azimuthal variation in the boundary conditions as experienced during a LOCA is discussed. The 2D framework utilized has been developed by the author of this report as part of his dissertation research ongoing at the University of South Carolina under the direction of Professor Travis W. Knight. The report briefly describes the 2D framework used and highlights in more detail the required developments in Bison to facilitate the extension of the axial relocation model to work within this framework. Finally, a few demonstration cases are provided that illustrate the benefit of using the new extended axial relocation model for the simulation of fuel rods under LOCA conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Experimental Study for Effects of Ballooning and Power Peak on a Coolability of Fuel Rod Bundle

For a large break loss-of-coolant accident (LB LOCA) in a pressurized water reactor, the clad temperature increases until the reflood phase. This causes ballooned fuel rods and fuel relocation, thereby reducing flow passage area of sub-channel and redistributing the flow and heat transfer in sub-channels. The fuel relocation increase the power at the ballooned region and temperature on the fuel rod. The research on the coolability of the deformed fuel and fuel relocation is very important for a prevention and mitigation of the severe accidents. In the present study, the reflood experiments were performed to investigate the coolability in the ballooned geometry and fuel relocation condition during the LB LOCA for Korean nuclear power plant, APR1400. The coolability experiments were carried out using three kinds of test sections. The first experiments using 6x6 intact rod bundle were performed as a base case. The second experiments simulated the coolability in the deformed fuel using 5x5 ballooned rod bundle. The last experiments were performed using another ballooned 5x5 rod bundle with local power increase to simulate fuel relocation. The fuel relocation phenomena increase the peak cladding temperatures compared with intact and ballooned only rod bundles.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Verification of the BISON fuel performance code

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Labo- ratory (USA) since 2009. The code is applicable to both steady and transient fuel behavior and is used to analyze 1D spherical, 2D axisymmetric, or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods and other well known fuel performance codes. Results from several assessment cases are reported, with emphasis on fuel centerline temperatures at various stages of fuel life, fission gas release, and clad deformation during pellet clad mechanical interaction (PCMI). BISON comparisons to fuel centerline temperature measurements are very good at beginning of life and reasonable at high burnup. Although limited to date, fission gas release comparisons are very good. Comparisons of rod diameter following significant power ramping are also good and demonstrate BISON’s unique ability to model discrete pellet behavior and accurately predict clad ridging from PCMI.

99 GENERAL AND MISCELLANEOUS↗
NASA NTRS1991

Advanced materials for space nuclear power systems

The overall philosophy of the research was to develop and characterize new high temperature power conversion and radiator materials and to provide spacecraft designers with material selection options and design information. Research on three candidate materials (carbide strengthened niobium alloy PWC-11 for fuel cladding, graphite fiber reinforced copper matrix composites for heat rejection fins, and tungsten fiber reinforced niobium matrix composites for fuel containment and structural supports considered for space power system applications is discussed. Each of these types of materials offers unique advantages for space power applications.

Titran, Robert H.↗
NASA NTRS1991

Advanced materials for space nuclear power systems

The overall philosophy of the research was to develop and characterize new high temperature power conversion and radiator materials and to provide spacecraft designers with material selection options and design information. Research on three candidate materials (carbide strengthened niobium alloy PWC-11 for fuel cladding, graphite fiber reinforced copper matrix composites for heat rejection fins, and tungsten fiber reinforced niobium matrix composites for fuel containment and structural supports) considered for space power system applications is discussed. Each of these types of materials offers unique advantages for space power applications.

Titran, Robert H.↗
DOE OSTI2024

Initial Assessment of CTF for Time-at-Temperature Applications

The US nuclear industry is interested in improving the economics of their fleet of light-water reactors (LWRs) by uprating US plants. One option being considered is to regain lost margin from overly conservative fuel safety limits. The current limit requires avoidance of critical heat flux (CHF) and prevents further operation of fuel that experiences a dry-out in boiling water reactors (BWRs) or departure from nucleate boiling (DNB) in pressurized water reactors (PWRs); however, it has been shown that temporary, mild dry-out of the fuel does not necessarily increase the risk of fuel failure during its normal anticipated operating life. Such mild dry-out or DNB events may occur during a plant anticipated operational occurrence (AOO), such as a locked rotor in a PWR or a pump trip in a BWR. The time-at-temperature (TAT) approach to regulating fuel operation aims to demonstrate that the fuel rod’s integrity is not challenged during such a mild transient that leads to CHF in which the fuel operates at an elevated temperature for a brief period of time. However, implementing this approach will require extensive fuel material experimental data, as well as supporting modeling and simulation (M&S) predictions, to ensure that the predicted fuel response during AOOs, with all applicable uncertainty considered, will not threaten the safety of the fuel during the transient or the remainder of its anticipated lifecycle. To address this need, a comprehensive effort is being proposed that includes generating cladding material data under TAT conditions, assessment of available code capabilities for TAT conditions, development of new mechanistic models, and demonstration of the M&S capabilities for AOOs of interest. This will require a joint effort between the Nuclear Energy Advanced Modeling and Simulation (NEAMS) and Advanced Fuels Campaign (AFC) programs, as well as close collaboration with nuclear industry stakeholders. The outcome of this collaboration will result in development and assessment of capabilities that can be used by the nuclear industry to support qualification of a TAT-based fuel failure criteria safety limit. This report focuses on the thermal hydraulics (T/H) modeling capabilities and summarizes currently available data for validating the T/H subchannel code CTF for TAT conditions, as well as preliminary assessment results of the code. The initial assessment also resulted in implementation of an alternative post-CHF heat transfer package, which has been shown to significantly improve accuracy. This report is not a final assessment and does not consider all available validation data; it is intended that a future assessment will more fully validate the code for this application.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

BISON validation of FeCrAl cladding mechanical failure during simulated reactivity-initiated accident conditions*

Here, a reactivity-initiated accident (RIA) is a postulated design basis accident in light water reactors (LWRs) in which a rapid reactivity insertion induces a fission rate increase and a fuel pellet temperature rise. During RIA, the fuel pellet thermally expands and may cause pellet-cladding mechanical interaction (PCMI). Separate effects PCMI tests were performed on C26M FeCrAl cladding tube samples, introducing biaxial stress via a well understood modified burst test (MBT) system. A high-speed camera in the MBT system captured the projections, covering a 360° view of the cladding deformation and enabling a digital image correlation (DIC) method to quantify the surface strains with high fidelity. Representative hot zero-power RIA mechanical loading conditions were applied to the sample, and the test duration ranged from 20 to 500 ms at an average temperature of 573 K. BISON finite element–based fuel performance code modeling was performed against the high-fidelity DIC data produced from the MBTs. Validation calculations were conducted with 2D models and systematically compared with test data of the burst time, burst pressure, burst hoop strains, and hoop strain rates. Based on the behaviors from the separate effects test, BISON calculations satisfactorily predicted the cladding deformation behaviors. Sensitivity analysis was conducted to identify highly influential mechanical properties responsible for the cladding failure behavior during the MBT experiments. The results highlight the significance of the cladding's mechanical strength in governing cladding strain, followed by the significance of the mechanical interactions between the pellet and the cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗