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DOE OSTI2021

Additive manufacturing of high-strength multiphase nanostructures in the binary Ni–Nb system for the discovery of new types of superalloys

Abstract Ni-based superalloys have been studied extensively due to their impressive mechanical properties, including strength and creep resistance at high temperatures. Growing interest surrounding additive manufacturing (AM) methods has led to recent investigations of alloys that are traditionally difficult to process, including Ni-based superalloys. Recent work has shown that AM methods enable high-throughput materials discovery and optimization of difficult- or impractical-to-process alloys, including those with high or even majority refractory element compositions. This work focuses on AM-enabled investigations of composition-dependent mechanical and microstructural properties for Ni–Nb binary alloys. Specifically, we report on the mechanical behavior of compositionally-graded Ni x Nb 1− x and uniform composition Ni 59.5 Nb 40.5 specimens made with AM. The AM fabrication process resulted in extraordinarily high strength, attributed to the formation of a dual-phase microstructure consisting of δ-Ni 3 Nb and µ-Ni 6 Nb 7 intermetallic compounds with nanostructured and multimodal grain size and eutectic lamellar spacing. Graphic Abstract

Jones, M. R.↗

DOE OSTI2013

Reducing the Cost of Thermal Energy Storage for Parabolic Trough Solar Power Plants

Thermal energy storage systems using phase change materials were evaluated for trough systems that use oil, steam, and high temperature salts as heat transfer fluids. A variety of eutectic salts and metal alloys were considered as phase change materials in a cascaded arrangement. Literature values of specific heat, latent heat, density, and other thermophysical properties were used in initial analyses. Testing laboratories were contracted to measure properties for candidate materials for comparison to the literature and for updating the models. A TRNSYS model from Phase 1 was further developed for optimizing the system, including a novel control algorithm. A concept for increasing the bulk thermal conductivity of the phase change system was developed using expanded metal sheets. Outside companies were contracted to design and cost systems using platecoil heat exchangers immersed in the phase change material. Laboratory evaluations of the one-dimensional and three-dimensional behavior of expanded metal sheets in a low conductivity medium were used to optimize the amount of thermal conductivity enhancement. The thermal energy storage systems were compared to baseline conventional systems. The best phase change system found in this project, which was for the high temperature plant, had a projected cost of $25.2 per kWhth, The best system also had a cost that was similar to the base case, a direct two-tank molten salt system.

14 SOLAR ENERGY↗

DOE OSTI2013

Modeling Solute Thermokinetics in LiCl-KCl Molten Salt for Nuclear Waste Separation (NEUP 10-418)

Recovery of actinides is an integral part of a closed nuclear fuel cycle. Pyrometallurgical nuclear fuel recycling processes have been developed in the past for recovering actinides from spent metallic and nitride fuels. The process is essentially to dissolve the spent fuel in a molten salt and then extract just the actinides for reuse in a reactor. Extraction is typically done through electrorefining, which involves electrochemical reduction of the dissolved actinides and plating onto a cathode. Knowledge of a number of basic thermokinetic properties of salts and salt-fuel mixtures is necessary for optimizing present and developing new approaches for pyrometallurgical waste processing. The properties of salt-fuel mixtures are presently being studied, but there are so many solutes and varying concentrations that direct experimental investigation is prohibitively time consuming and expensive (particularly for radioactive elements like Pu). Therefore, there is a need to reduce the number of required experiments through modeling of salt and salt-fuel mixture properties. This project will develop first-principles-based molecular modeling and simulation approaches to predict fundamental thermokinetic properties of dissolved actinides and fission products in molten salts. The focus of the proposed work is on property changes with higher concentrations (up to 5 mol%) of dissolved fuel components, where there is still very limited experimental data. The properties predicted with the modeling will be density, which is used to assess the amount of dissolved material in the salt; diffusion coefficients, which can control rates of material transport during separation; and solute activity, which determines total solubility and reduction potentials used during electrorefining. The work will focus on La, Sr, and U, which are chosen to include the important distinct categories of lanthanides, alkali earths, and actinides, respectively. Studies will be performed using LiCl-KCl salt at the eutectic composition (58 mol% LiCl, 42 mol% KCl), which is used for treating spent EBR-II fuel. The same process being used for EBRII fuel is currently being studied for widespread international implementation. The methods will focus on first-principles and first- principles derived interatomic potential based simulations, primarily using molecular dynamics. Results will be validated against existing literature and parallel ongoing experimental efforts. The simulation results will be of value for interpreting experimental results, validating analytical models, and for optimizing waste separation by potentially developing new salt configurations and operating conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

A Brief Review of Past INL Work Assessing Radionuclide Content in TMI-2 Melted Fuel Debris: The Use of 144 Ce as a Surrogate for Pu Accountancy

This report serves as a literature review of prior work performed at Idaho National Laboratory, and its predecessor organizations Idaho National Engineering Laboratory (INEL) and Idaho National Engineering and Environmental Laboratory (INEEL), studying radionuclide partitioning within the melted fuel debris of the reactor of the Three Mile Island 2 (TMI-2) nuclear power plant. The purpose of this review is to document prior published work that provides supporting evidence of the utility of using 144 Ce as a surrogate for plutonium within melted fuel debris. When the TMI-2 accident occurred no quantitative nondestructive analysis (NDA) techniques existed that could assay plutonium in the unconventional wastes from the reactor. However, unpublished work performed at INL by D. W. Akers in the late 1980s through the 1990s demonstrated that passive gamma-ray spectrometry of 144 Ce could potentially be used to develop a semi-quantitative correlation for estimating plutonium content in these materials. The fate and transport of radioisotopes in fuel from different regions of the core, including uranium, fission products, and actinides, appear to be well characterized based on the maximum temperature reached by fuel in different parts of the core and the melting point, boiling point, and volatility of those radioisotopes. Also, the chemical interactions between fuel, fuel cladding, control elements, and core structural components appears to have played a large role in determining when and how fuel relocation occurred in the core; perhaps the most important of these reaction appears to be related to the formation of mixed-material alloys, eutectics, in the fuel cladding. Because of its high melting point, low volatility, and similar chemical behavior to plutonium, the element cerium appears to have behaved similarly to plutonium during the evolution of the TMI-2 accident. Anecdotal evidence extrapolated from open-source literature strengthens this logical feasibility for using cerium, which is rather easy to analyze using passive nondestructive analysis gamma-ray spectrometry, as a surrogate for plutonium in the final analysis of TMI-2 melted fuel debris. The generation of this report is motivated by the need to perform nuclear material accountancy measurements on the melted fuel debris that will be excavated from the damaged nuclear reactors at the Fukushima Daiichi nuclear power plant in Japan, which were destroyed by the Tohoku earthquake and tsunami on March 11, 2011. Lessons may be taken from prior U.S. work related to the study of the TMI-2 core debris to support the development of new assay methods for use at Fukushima Daiichi. While significant differences exist between the two reactor systems (pressurized water reactor (TMI-2) versus boiling water reactor (FD), fresh water post-accident cooing (TMI-2) versus salt water (FD), maintained containment (TMI-2) versus loss of containment (FD)) there remain sufficient similarities to motivate these comparisons.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2013

Control of a Uranium-Hydride Reactor with Deuterium-Hydrogen Exchange

A 70 MWt reactor is evaluated that utilizes uranium-hydride fuel and is cooled by lead-bismuth eutectic (LBE). The concept utilizes a high-density, high-density-fuel outer annulus which surrounds a lowdensity, powder-fuel central core. Prior to reactor startup, the entire core is initially saturated with deuterium (UD3). To start the reactor, hydrogen flows through the central, low density powder (with protons gradually replacing deuterons) until criticality is reached. Over the remaining lifetime of the reactor, with the goal of 10 years, hydrogen is continually added to the outer, high-density core to compensate for burnup reactivity loss. Throughout lifetime, the inner region serves as a standard control/shutdown rod, with the ability to replace the H with D relatively quickly (because of the low density, high surface area of the fuel). The scope of this work is to determine if a reactor could be designed to meet the above guidelines and if possible estimate the size of the smallest reactor that could meet the guidelines.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

The Nitrogen-Nitride Anode.

Nitrogen gas N 2 can be reduced to nitride N -3 in molten LiCl-KCl eutectic salt electrolyte. However, the direct oxidation of N -3 back to N 2 is kinetically slow and only occurs at high overvoltage. The overvoltage for N -3 oxidation can be eliminated by coordinating the N -3 with BN to form the dinitridoborate (BN 2 -3 ) anion which forms a 1-D conjugated linear inorganic polymer with -Li-N-B-N- repeating units. This polymer precipitates out of solution as Li 3 BN 2 which becomes a metallic conductor upon delithiation. Li 3 BN 2 is oxidized to Li + + N 2 + BN at about the N 2 /N -3 redox potential with very little overvoltage. In this report we evaluate the N 2 /N -3 redox couple as a battery anode for energy storage.

Delnick, Frank M.↗

DOE OSTI2015

Concept Feasibility Report for Electroplating Zirconium onto Uranium Foil - Year 2

The Fuel Fabrication Capability within the U.S. High Performance Research Reactor Conversion Program is funded through the National Nuclear Security Administration (NNSA) NA-26 (Office of Material Management and Minimization). An investigation was commissioned to determine the feasibility of using electroplating techniques to apply a coating of zirconium onto depleted uranium/molybdenum alloy (U-10Mo). Electroplating would provide an alternative method to the existing process of hot roll-bonding zirconium foil onto the U-10Mo fuel foil during the fabrication of fuel elements for high-performance research reactors. The objective of this research was to develop a reproducible and scalable plating process that will produce a uniform, 25 μm thick zirconium metal coating on U-10Mo foil. In previous work, Pacific Northwest National Laboratory (PNNL) established a molten salt electroplating apparatus and protocol to plate zirconium metal onto molybdenum foil (Coffey 2015). During this second year of the research, PNNL furthered this work by moving to the U-10Mo alloy system (90 percent uranium:10 percent molybdenum). The original plating apparatus was disassembled and re-assembled in a laboratory capable of handling low-level radioactive materials. Initially, the work followed the previous year’s approach, and the salt bath composition was targeted at the eutectic composition (LiF:NaF:ZrF4 = 26:37:37 mol%). Early results indicated that the formation of uranium fluoride compounds would be problematic. Other salt bath compositions were investigated in order to eliminate the uranium fluoride production (LiF:NaF = 61:39 mol% and LiF:NaF:KF = 46.5:11.5:42 mol% ). Zirconium metal was used as the crucible for the molten salt. Three plating methods were used—isopotential, galvano static, and pulsed plating. The molten salt method for zirconium metal application provided high-quality plating on molybdenum in PNNL’s previous work. A key advantage of this approach is that plating can be performed under conditions that would greatly reduce the quantity of intermetallics that form at the interface between the zirconium and U-10Mo; unlike roll bonding, the molten salt plating approach would allow for complete coverage of the U-10Mo foil with zirconium. When utilizing the experimental parameters developed for zirconium plating onto molybdenum, a uranium fluoride reaction product was formed at the Zr/U-10Mo interface. By controlling the initial plating potential, the uranium fluoride could be prevented; however, the targeted zirconium thickness (25 ±12.5 μm) could not be achieved while maintaining 100% coverage.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Oak Ridge National Laboratory Analysis of Waste Isolation Pilot Plant Samples: Integrated Summary Report

Nondestructive and destructive analytical methods were used to analyze samples of debris collected from the top of R15 C5, the lip of R16 C4, the parent drum, and the continuous air monitor (CAM) and fixed air sampler (FAS) filters in addition to MgO samples collected from the tops of R14 C2, R14 C4, R14 C6 and R16 C4. The analytical evidence supports the conclusion that an exothermic oxidation reaction of Swheat Scoop® by metal nitrate salts produced sufficient quantities of gas to breach drum 68660. The metals and the acidic environment found in drum 68660 are known to accelerate these oxidation reactions. Moreover, the ratio of metal nitrate salts in drum 68660 was in a range in which a low melting eutectic mixture could form, which would also accelerate the oxidation reactions. Nitrated organics were not detected and they are not necessary to produce the exothermic reaction leading to the breach of drum 68660. All the contents of the drum were not consumed and the material ejected from drum 68660 either did not experience temperatures over 200 °C, or their exposure to elevated temperatures was very limited in duration (<seconds).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Final Report on MEGAPIE Target Irradiation and Post-Irradiation Examination

Megawatt pilot experiment (MEGAPIE) was successfully performed in 2006. One of the important goals of MEGAPIE is to understand the behaviour of structural materials of the target components exposed to high fluxes of high-energy protons and spallation neutrons in flowing LBE (liquid lead-bismuth eutectic) environment by conducting post-irradiation examination (PIE). The PIE includes four major parts: non-destructive test, radiochemical analysis of production and distribution of radionuclides produced by spallation reaction in LBE, analysis of LBE corrosion effects on structural materials, T91 and SS 316L steels, and mechanical testing of the T91 and SS 316L steels irradiated in the lower part of the target. The non-destructive test (NDT) including visual inspection and ultrasonic measurement was performed in the proton beam window area of the T91 calotte of the LBE container, the most intensively irradiated part of the MEGAPIE target. The visual inspection showed no visible failure and the ultrasonic measurement demonstrated no detectable change in thickness in the beam window area. Gamma mapping was also performed in the proton beam window area of the AlMg 3 safety-container. The gamma mapping results were used to evaluate the accumulated proton fluence distribution profile, the input data for determining irradiation parameters. Radiochemical analysis of radionuclides produced by spallation reaction in LBE is to improve the understanding of the production and distribution of radionuclides in the target. The results demonstrate that the radionuclides of noble metals, 207 Bi, 194 Hg/Au are rather homogeneously distributed within the target, while radionuclides of electropositive elements are found to be deposited on the steel-LBE interface. The corrosion effect of LBE on the structural components under intensive irradiation was investigated by metallography. The results show that no evident corrosion damages. However, unexpected deep cracks were found in the EBW (electron beam weld) of the LBE container in the intensive irradiation zone of the target, which should be formed during irradiation. In the SS 316L steel of the flow guide tube, inclusions or precipitates enriched with O, Si, S, Ca, Ti and Mn were observed. Many of them are very long, up to a few mm, and located on grain boundaries along the extrusion direction of the tube. The degradation of the mechanical properties of the T91 and SS 316L steels has been investigated by conducting tensile tests on the specimens extracted from the T91 and SS 316L components in the intensive irradiation region. The results obtained from the proton beam window of the T91 calotte exhibit a good ductility of T91 steel after irradiation at 6-7 dpa (displacement per atom) in contact with flowing LBE.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Thermal Testing and Quality Assurance of BGA LCC & QFN Electronic Packages

The purpose of this project is to experimentally validate the thermal fatigue life of solder interconnects for a variety of surface mount electronic packages. Over the years, there has been a significant amount of research and analysis in the fracture of solder joints on printed circuit boards. Solder is important in the mechanical and electronic functionality of the component. It is important throughout the life of the product that the solder remains crack and fracture free. The specific type of solder used in this experiment is a 63Sn37Pb eutectic alloy. Each package was surrounded conformal coating or underfill material.

36 MATERIALS SCIENCE↗

DOE OSTI2017

Final report for SERDP WP-2209 Replacement melt-castable formulations for Composition B

During this project we investigated a number of energetic materials both old and new and determined that most of them were unsuitable due to safety or sensitivity reasons. Unsuccessful coformulants include TNAZ and BNFF for volatility reasons, and DAAF due to thermal compatibility issues. The powerful explosive HMX became a focus of the work in later stages as it conferred excellent power while being commonly available in well-regulated particle size lots and is chemically compatible in the melt with many coformulants. Ultimately three preferred formulations emerged from this work: a formulation tested on large scale by ARDEC involving PrNQ and HMX; a formulation tested at ARDEC and LANL using a nitrate salt eutectic and HMX; a formulation tested at LANL using LLM-201 and HMX.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2017

Modeling cookoff of HMX based PBX explosives

We have previously developed a PBX 9501 cookoff model for the plastic bonded explosive PBX 9501 consisting of 95 wt% octahydro-1,3,5,7-tetranitro-1,3,5,7-tetrazoncine (HMX), 2.5 wt% Estane® 5703 (a polyurethane thermoplastic), and 2.5 wt% of a nitroplasticizer (NP): BDNPA/F, a 50/50 wt% eutectic mixture bis(2,2-dinitropropyl)-acetal (BDNPA) and bis(2,2-dinitropropyl)-formal (BDNPF). This fivestep model includes desorption of water, decomposition of the NP to form NO 2 , reaction of the NO 2 with Estane and HMX, and decomposition of HMX [1]. This model has been successfully validated with data from six laboratories with scales ranging from 2 g to more than 2.5 kg of explosive. We have determined, that the PBX 9501 model can be used to predict cookoff of other plastic bonded explosives containing HMX and an inert binder, such as LX-04 consisting of 85 wt% HMX and 15 wt% Viton A (vinylidine fluoride/hexafluoropropylene copolymer), LX-07 (90 wt% HMX and 10 wt% Viton A), LX- 10-0 (95 wt% HMX and 5 wt% Viton A), and LX-14 consisting of 95.5 wt % HMX and 4.5 wt% Estane® 5702-F1 (a polyurethane thermoplastic). Normally our cookoff models are verified using Sandia’s Instrumented Thermal Initiation (SITI) experiment. However, SITI data for LX-04, LX-07, LX-10-0, and LX-14 are not available at pressed density; although, some molding powder SITI data on LX-10-0 and LX-14 exists. Tarver and Tran [2] provide some one-dimensional time-to-explosion (ODTX) data for these explosives. The applicability of the PBX 9501 model to LX-04, LX-07, LX-10-0, AND LX-14 was made using this ODTX data [2]. The PBX 9501 model is applied to these other explosives by accounting for the correct amount of HMX in the explosive and limiting the NP reaction. We have found the PBX 9501 model to be useful for predicting the response of these PBXs to abnormal thermal environments such as fire.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2017

Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate constant (3 times and 10 times that of the rate constant for APMT) had a negligible impact on the early stages of the accident and minor impacts on the accident progression after the first relocation of the fuel. At temperatures below 1,500°C, increasing the rate constant for APMT by a factor of 10 still resulted in only minor FeCrAl oxidation. In general, the gains afforded by the FeCrAl enhanced ATF concept with respect to accident sequence timing and combustible gas generation are consistent with previous efforts. Compared with the traditional Zircaloy-based cladding and channel box system, the FeCrAl concept could provide a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. For example, a station blackout was simulated in which cooling water injection was lost 36 hours after shutdown. The timing to first fuel relocation was delayed by approximately 5 h for the FeCrAl ATF concept compared with that of the traditional Zircaloy-based cladding and channel box system.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Rare Earth Electrochemical Property Measurements and Phase Diagram Development in a Complex Molten Salt Mixture for Molten Salt Recycle

Pyroprocessing is a promising alternative for the reprocessing of used nuclear fuel (UNF) that uses electrochemical methods. Compared to the hydrometallurgical reprocessing method, pyroprocessing has many advantages such as reduced volume of radioactive waste, simple waste processing, ability to treat refractory material, and compatibility with fast reactor fuel recycle. The key steps of the process are the electro-refining of the spent metallic fuel in the LiCl-KCl eutectic salt, which can be integrated with an electrolytic reduction step for the reprocessing of spent oxide fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Advancing Coal Catalytic Gasification to Promote Optimum Syngas Production

An investigation was pursued to improve the use of low rank sub-bituminous coal by using suitable catalysts to advance gasification and create cleaner synthesis gas. Bench- and pilot-scale experiments were performed to measure key reactive properties and verify the use of red mud catalyst. The experiments provided reaction data for the development of kinetic models and validation data for CFD simulations. The reaction kinetics were analyzed to provide an understanding of the major pathways involved and identify the necessary reaction mechanisms needed for the computational models. The reaction kinetics model was implemented into the computational fluid dynamics (CFD) code MFiX to simulate, predict and validate the chemistry model with the experiments. It was shown that red mud as a catalyst lowered methane formation during coal gasification. Additional benefits were that red mud destroyed tars formation, promoted water gas shift reaction resulting in a syngas rich in hydrogen. In the presence of red mud catalyst, biomass feedstocks high in ash did not form clinker in the fluidized bed because it prevented the formation of eutectic mixtures. The reaction kinetics model was reasonably robust and the CFD predictions using the model were in good agreement with the experiments.

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2018

Phenomena Important in Liquid Metal Reactor Simulations

The U.S. Nuclear Regulatory Commission (NRC) is preparing for the future licensing of advanced reactors that will be very different from current light water reactors. Part of the NRC preparation strategy is to identify the simulation tools that will be used for confirmatory safety analysis which includes normal operation and abnormal conditions. This report advances that strategy for reactors with a fast neutron spectrum that use liquid metal coolants. This includes reactors using sodium, lead, or a lead-bismuth eutectic. Although all types are discussed in this report, the emphasis is on simulating sodium-cooled reactors as more information is available on those designs. The specific designs discussed in the report are a subset of many deigns being considered in the U.S. and elsewhere but they are considered the most likely to submit information to the NRC in the near future. The objective herein is to identify the dominant physical phenomena and modeling requirements expected for the simulation of liquid metal fast spectrum reactors. Phenomena are discussed in terms of normal and transient/accident conditions, and the corresponding modeling tools for neutronics, heat transfer, and fluid dynamics. The study makes use of the experience with such concepts, for example, from the Experimental Breeder Reactor (EBR-II) that ran for almost 30 years at Idaho National Laboratory; from recent conceptual design studies such as for the PRISM concept; and from similar technology-gap studies carried out recently for the Department of Energy. Lists of important phenomena were generated for normal operation and transients/accidents that will need to be simulated. These lists can be used by a panel of experts as the starting point for generating a phenomena identification and ranking table (PIRT). The PIRT exercise would determine what research and development may be required before the necessary simulation tools are available.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Design and Thermal Analysis for Irradiation of Silicon Carbide Joint Specimens in the High Flux Isotope Reactor

This report provides a summary of the irradiation vehicle design and thermal analysis of SiC joint specimens planned for irradiation in the flux trap of the High Flux Isotope Reactor (HFIR). Two different capsule designs will be used to accommodate the two different specimen geometries: a small torsion joint specimen geometry to measure mechanical and thermal properties, and a tube and end plug joint representative cladding geometry to demonstrate strength and hermeticity retention. The capsule designs, with target temperatures of 350° and 750°C, will accommodate either sixteen torsion joint specimens or one tube end plug joint specimen. Three joint variations will be studied in each capsule design: a hybrid SiC (preceramic polymer with chemical vapor deposition (CVD) SiC), a transient eutectic phase (TEP) process, and an oxide process.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Design and analysis of oxidation tests to inform FeCrAl ATF severe accident models

36 MATERIALS SCIENCE↗