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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 91 records · Page 5

SAS4A/SASSYS-1 Commercial Grade Dedication Example Report for a Generic Sodium Pool-Type Fast Reactor Application

In the U.S., a key component of the commercialization of advanced reactors is completion of a license application, which must ultimately be approved by the Nuclear Regulatory Commission (NRC). The approval of the license application by the NRC is contingent on satisfactory demonstration of the design basis and the response of the advanced reactor design to transient and accident scenarios using accepted codes and methods. This report describes the qualification and dedication requirements that the advanced reactor safety analysis system software SAS4A/SASSYS-1 are expected to need to fulfill to be used for sodium-cooled pool-type fast reactor licensing. The qualification and dedication requirements are identified through performance critical characteristics and evaluation model acceptance criteria representative of the advanced reactor design considered for licensing. This document captures, additionally, the verification process developed to demonstrate that the software fulfills the qualification and dedication requirements for a generic sodium-cooled pool-type fast reactor as part of the commercial grade dedication process. Like most software that has primarily existed in the research and development space, the most significant challenge facing SAS4A/SASSYS-1 for use in a licensing framework is the availability of a documentation basis describing the code pedigree. SAS4A/SASSYS-1 has been used for licensing of the fast flux test facility (FFTF) and the JOYO sodium-cooled fast reactor in Japan, as well as the design of the CRBR Plant. However, the historical verification and validation (V&V) activities supporting SAS4A/SASSYS-1 development do not align with modern software quality assurance (SQA) and V&V requirements. Two approaches to use of SAS4A/SASSYS-1 in a commercial licensing framework have been identified: commercial-grade dedication (CGD) and software qualification. The methods and requirements prescribed in the ASME NQA-1-2008/2009 Standard and Regulatory Guide 1.203 on the evaluation model development and assessment process (EMDAP) have been used as guidance to define the CGD and qualification processes, respectively. A qualification and dedication requirements matrix has been developed which utilizes fundamental software verification. In this process, software verification is defined as a software quality process aimed at defining software requirement specifications, developing software design documentation, and performing and documenting acceptance testing of the code against requirements. A key element of software qualification and dedication includes determination of software acceptance with respect to critical characteristics relevant to the functional requirements of the software. To assist with identification of cross-cutting transient phenomena and functional requirements, domestic SFR vendor designs have been reviewed to identify a reference SFR design. For this report, the reference design is defined as a pool-type reactor with metal alloy fuel, a liquid-metal intermediate heat transport system, and passive decay heat rejection systems. Given this reference, a series of high-level cross-cutting phenomena was identified for a general class of single-fault undercooling or reactivity insertion transients that scopes the design basis space, with the goal of assisting with prioritization of documentation development efforts for key transient models in SAS: 1) Reactivity feedback response prior to scram; 2) System-wide thermal inertia; 3) Transition in natural circulation flow regime in heat removal systems; 4) Decay heat generation; 5) Steady-state fuel characterization; 5) Clad/fuel behavior at elevated temperatures; 6) Point kinetics and decay heat; 7) Pump coastdown behavior; 8) Core flow redistribution in loss of forced convection; 9) Pool stratification. As a demonstration of CGD of SAS4A/SASSYS-1 for a sodium pool reactor, a software qualification and dedication gap analysis as it relates to code documentation has been performed. This effort leverages the framework established as part of the SAS4A/SASSYS-1 SQA Program. This CGD demonstration provides a framework that vendors can build upon to demonstrate the applicability of the SAS4A/SASSYS-1 software for licensing a sodium-cooled pool-type fast reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Fracture control of H-O engine components

An investigation was made to obtain the material characterization and fatigue crack propagation data necessary to establish the salient characteristics of a Ti-6Al-2.5Sn(ELI) alloy fuel pump impeller to be used in a cryogenic service environment. Testing variables considered were: coupon orientation, frequency, load range ratio, and temperature. Data analysis correlated crack propagation data from conventional laboratory coupons with data from a parallel sided rotating disk used to model rotor stresses. Four major design recommendations when bore regions of fuel pump impellers to be operated in cryogenic environments are to be relatively highly stressed are discussed.

Ryder, J. T.↗

Fracture Growth Testing of Titanium 6AL-4V in AF-M315E

The Green Propellant Infusion Mission (GPIM) will demonstrate the performance of AF-M315E monopropellant in orbit. Flight certification requires a safe-life analysis of the titanium alloy fuel tank to ensure inherent flaws will not cause failure during the design life. Material property inputs for this analysis require testing to determine the stress intensity factor for environmentally-assisted cracking (K (sub EAC)) of Ti 6Al-4V in combination with the AF-M315E monopropellant. Testing of single-edge notched specimens SE(B) representing the bulk tank membrane and weld material were performed in accordance with ASTM E1681. Specimens with fatigue pre-cracks were loaded into test fixtures so that the crack tips were exposed to the monopropellant at 50 degrees Centigrade for a duration of 1,000 hours. Specimens that did not fail during exposure were opened to inspect the crack surfaces for evidence of crack growth. The threshold stress intensity value, KEAC, is the highest applied stress intensity that produced neither a failure of the specimen during the exposure nor showed evidence of crack growth. The threshold stress intensity factor of the Ti 6Al-4V forged tank material when exposed to AF-M315E monopropellant was found to be at least 22.0 kilopounds per square inch. The stress intensity factor of the weld material was at least 31.3 kilopounds per square inch.

AF-M315E↗

Fracture Mechanics Testing of Titanium 6AL-4V in AF-M315E

The Green Propellant Infusion Mission (GPIM) will demonstrate the performance of AF-M315E monopropellant on orbit. Flight certification requires a safe-life analysis of the titanium alloy fuel tank to ensure inherent processing flaws will not cause failure during the design life of the tank. Material property inputs for this analysis require testing to determine the stress intensity factor for environment-assisted cracking (KEAC) of Ti 6Al-4V in combination with the AF-M315E monopropellant. Testing of single-edge notched, or SE(B), specimens representing the bulk tank membrane and weld material were performed in accordance with ASTM E1681. Specimens with fatigue pre-cracks were loaded into test fixtures so that the crack tips were exposed to AF-M315E at 50 C for a duration of 1,000 hours. Specimens that did not fail during exposure were opened to inspect the crack surfaces for evidence of crack growth. The threshold stress intensity value, KEAC, is the highest applied stress intensity that produced neither a failure of the specimen during the exposure nor showed evidence of crack growth. The threshold stress intensity factor for environment-assisted cracking of the Ti 6Al-4V forged tank material was found to be at least 22 ksi√in and at least 31 ksi√in for the weld material when exposed to AF-M315E monopropellant.

AF-M315E↗

Assessment of Measurement Uncertainties in the Jupiter High-240 Experiment

The Jupiter High-240 experiment performed in May of 2019 was previously discussed as a variant of the original Jupiter experiment incorporating plutonium metal alloy fuel plates with higher 240Pu content and lead plates, using both a reference configuration and a second configuration where eight lead plates were replaced with aluminum to simulate voiding. Measurements were recorded for experiment period, the “pressure” of the Comet ram upon closure for each near-critical measurement, and temperature. The experiment reactor period is the time it would take to increase the neutron population by a factor of e. For this experiment, the copper reflectors and upper third of the fuel sits upon a support structure with the lower fuel arrays raised up into the center of the reflectors using a ram (see Fig. 1). The recorded logbook temperature for each measurement corresponds to a resistance temperature detector (RTD) located at the top center of the upper fuel array. This paper summarizes the evaluated uncertainties for the Jupiter High 240 experiment as contributed via the recorded measurements and nuclear data and their assessed impact upon the computation of system reactivity and eigenvalue.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Role of plastic deformation in wear of copper and copper - 10-percent-aluminum alloy in cryogenic fuels

High-purity copper specimens and a copper-aluminum (10%) alloy specimen were subjected to sliding against Type 440 C in cryogenic fuel environments. It was found that virtually all wear occurred by the plastic deformation of a recrystallized layer extending to about 10 micrometers below the wear scar surface of the copper or copper alloy. The wear debris was in the form of a layered structure adhering to the exit region of the wear scar. Measurements on the high purity copper specimens indicated that the wear rate was proportional to the applied load and to the sliding velocity squared. A physical model of the wear process is proposed to account for these observations.

Bill, R. C.↗

A new high strength alloy for hydrogen fueled propulsion systems

This paper describes the development of a high-strength alloy (1241 MPa ultimate and 1103 MPa yield, with little or no degradation in hydrogen) for application in advanced hydrogen-fueled rocket engines. Various compositions of the Fe-Ni-Co-Cr system with elemental additions of Cb, Ti and Al are discussed. After processing, notched tensile specimens were tested in 34.5-MPa hydrogen at room temperature, as the main screening test. The H2/air notch tensile ratio was used as the selection/rejection criterion. The most promising alloys are discussed.

Mcpherson, W. B.↗

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↗

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↗

Alloys for inert matrix fuel compositions, and methods of making the same

In one embodiment, an alloy includes: Zr; Fe; Cu; Ta in an amount from about 1 wt % to about 3 wt %; and one or more optional constituents selected from: Ti, Be, and Nb; and wherein the alloy comprises a ductile phase and a nanoprecipitate hard phase. According to another embodiment, a method of forming an inert matrix nuclear fuel includes: packing a hollow structure with fuel pellets and alloy precursor pellets; heating the fuel pellets and the alloy precursor pellets to at least a melting temperature of an alloy to be formed by melting the alloy precursor pellets; and solidifying the alloy into a matrix surrounding the fuel pellets. The alloy precursor pellets independently comprise: Zr; Fe; Cu; Ta present in an amount from about 1 to about 3 wt %; and one or more optional alloy constituents selected from: Ti, Be, and Nb.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Hydrothermal Corrosion of Latest Generation of FeCrAl Alloys for Nuclear Fuel Cladding

After the Fukushima nuclear disaster, the nuclear materials community has been vastly investing in accident tolerant fuel (ATF) concepts to modify/replace Zircaloy cladding material. Iron–chromium–aluminum (FeCrAl) alloys are one of the leading contenders in this race. In this study, we investigated FA-SMT (or APMT-2), PM-C26M, and Fe17Cr5.5Al over a time period of 6 months in simulated BWR environments and compared their performance with standard Zirc-2 and SS316 materials. Our results implied that water chemistry along with alloy chemistry has a profound effect on the corrosion rate of FeCrAl alloys. Apart from SS316 and Zirc-2 tube specimens, all FeCrAl alloys showed a mass loss in hydrogen water chemistry (HWC). FA-SMT displayed minimal mass loss compared to PM-C26M and Fe17Cr5.5Al because of its higher Cr content. The mass gain of FeCrAl alloys in normal water chemistry (NWC) is significantly less when compared to Zirc-2.

36 MATERIALS SCIENCE↗

Fabrication and Characterization of Candidate Alloys for Advanced LEU Fuel Concepts

A set of novel uranium-based alloys has been proposed for use in advanced LEU fuel systems. This study seeks to develop methods to fabricate single-phase alloys and evaluate the compatibilities of the alloys for use in composite fuel systems. The as-cast microstructure of each alloy was characterized via scanning electron microscopy (SEM), and candidate alloys were further down-selected for homogenization and interaction testing. The U 22 Zr 20 Nb 58 alloy was successfully fabricated as a single-phase system, as demonstrated via X-ray diffraction (XRD) and SEM. The thermal and mechanical properties of the U-Nb-Zr alloy were further assessed via differential scanning calorimetry (DSC), laser flash analysis (LFA), and Vickers’ micro-hardness testing. This alloy also demonstrated improved compatibility with fuel candidate alloys from conventional matrix materials and may be a viable option for use in advanced fuel systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A Model of Hydrogen Solubility in Palladium-Silver Alloys

Fusion fuel cycle designs depend on palladium and palladium-silver alloys. The selective permeability of these is used to separate hydrogen isotopes from other elements. Predicting the solubility of hydrogen isotopes in palladium-silver alloys is important for the design of key unit operations for tritium processing. This paper presents a model of hydrogen isotope solubility in palladium and palladium-silver alloys based on Guggenheim’s quasichemical lattice theory. The model parameters were determined as functions of isotope molecular weight and the silver content of the alloy. The model predicts reasonable values for the solubility of protium, deuterium, and tritium over a wide range of temperatures, pressures, and alloy compositions.

08 HYDROGEN↗

An Ab Initio Molecular Dynamics Study of Key Thermodynamic Input Parameters for Computer Simulation of U-6Nb Solidification

The key to metallic fuel development is the fabrication of uranium metal and alloys into fuel forms. U-Nb alloys are one of the best candidates for a metallic fuel alloy with high-temperature strength sufficient to support the core, acceptable nuclear properties, good fabricability, and compatibility with usable coolant media. Melt processing has been a key component of the metallic fuel cycle, and process models require thermophysical parameters at elevated temperatures, particularly above the melting temperatures, regarding which experimental data are scarce, for accurate simulations and process development. By means of ab initio density-functional theory (DFT) quantum molecular dynamics (QMD), we have calculated the main thermophysical parameters—the density, thermal expansion coefficient, specific heat, thermal conductivity, melting temperature, latent heat of fusion, and viscosity—used in the modeling of the U-6 wt.% Nb alloy casting. The melting temperature of the U-6 wt.% Nb alloy at ambient pressure is obtained by means of QMD simulations using the Z-method. The ambient volume change and latent heat of melting of U-6 wt.% Nb are also derived from QMD simulations in conjunction with analytical fitting for the energy and pressure. The thermal conductivity for the solid U-Nb alloy is calculated from the semi-classical Boltzmann transport equation combined with an estimate of the electron relaxation time obtained from DFT simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Corrosion of Ferrous Structural Alloys in Biomass Derived Fuels and Organic Acids

This work investigated the corrosivity of biomass-derived liquid fuels and organic acids associated with bio-oils and conducted a corrosion compatibility evaluation of several structural ferrous alloys in the fuels and acids. Here, the acidity parameters and solution conductivity values were higher in raw pyrolysis bio-oils than postprocessed oils with significant corrosion attack for carbon steel and Fe-2.25Cr-1Mo. However, higher-grade stainless steels, ferritic Fe-15Cr (15Cb) and austenitic types 304L and 316L, showed negligible corrosion loss, whereas lower-Cr ferritic grades 409, 410, and 416 were attacked. Metal leaching tests and electrochemical impedance spectroscopy measurements revealed superior corrosion resistance of 316 over 410 stainless steels in solutions of formic and oxalic acids that can be present in bio-oils. The corrosion evaluation methods utilized in this work provide effective screening tools to identify the compatible structural alloys for bio-oils.

09 BIOMASS FUELS↗