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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 1,027 records · Page 57
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Identification of nuclear effects in neutrino-carbon interactions at low three-momentum transfer

Two different nuclear-medium effects are isolated using a low three-momentum transfer subsample of neutrino-carbon scattering data from the MINERvA neutrino experiment. The observed hadronic energy in charged-current νμ interactions is combined with muon kinematics to permit separation of the quasielastic and Δ(1232) resonance processes. First, we observe a small cross section at very low energy transfer that matches the expected screening effect of long-range nucleon correlations. Second, additions to the event rate in the kinematic region between the quasielastic and Δ resonance processes are needed to describe the data. The data in this kinematic region also have an enhanced population of multiproton final states. Contributions predicted for scattering from a nucleon pair have both properties; the model tested in this analysis is a significant improvement but does not fully describe the data. We present the results as a double-differential cross section to enable further investigation of nuclear models. Furthermore, improved description of the effects of the nuclear environment are required by current and future neutrino oscillation experiments.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2016

Development and tests of molybdenum armored copper components for MITICA ion source

In order to prevent detrimental material erosion of components impinged by back-streaming positive D or H ions in the megavolt ITER injector and concept advancement beam source, a solution based on explosion bonding technique has been identified for producing a 1 mm thick molybdenum armour layer on copper substrate, compatible with ITER requirements. Prototypes have been recently manufactured and tested in the high heat flux test facility Garching Large Divertor Sample Test Facility (GLADIS) to check the capability of the molybdenum-copper interface to withstand several thermal shock cycles at high power density. This paper presents both the numerical fluid-dynamic analyses of the prototypes simulating the test conditions in GLADIS as well as the experimental results.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Benchmark Specifications for Select Experiments Conducted at the Kansas State University Gallium Thermal-hydraulic Experiment Facility

The Department of Energy (DOE) – Nuclear Energy University Programs (NEUP) supported the creation and operation of the Gallium Thermal-hydraulic Experiment (GaTE) facility at Kansas State University (KSU) as part of a larger effort to understand thermal stratification behavior in liquid-metal-cooled reactors. GaTE was designed to simulate transients in a reactor plenum that are known to cause thermal stratification. High-reliability and high-resolution measurements describing stratification behavior in the coolant were collected for use as experimental benchmarks in validation efforts for computational models. The results of these tests contribute to a greater understanding of thermal stratification behavior of liquid metal under various configurations and operating conditions. This report provides a complete description of the benchmark problem, including all necessary details and description of a set of four forced flow and four natural circulation tests and measured data for comparison with model results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2003

Simulation, Model Verification and Controls Development of Brayton Cycle PM Alternator: Testing and Simulation of 2 KW PM Generator with Diode Bridge Output

Professor Stankovic will be developing and refining Simulink based models of the PM alternator and comparing the simulation results with experimental measurements taken from the unit. Her first task is to validate the models using the experimental data. Her next task is to develop alternative control techniques for the application of the Brayton Cycle PM Alternator in a nuclear electric propulsion vehicle. The control techniques will be first simulated using the validated models then tried experimentally with hardware available at NASA. Testing and simulation of a 2KW PM synchronous generator with diode bridge output is described. The parameters of a synchronous PM generator have been measured and used in simulation. Test procedures have been developed to verify the PM generator model with diode bridge output. Experimental and simulation results are in excellent agreement.

Anna V Stankovic↗
DOE OSTI2016

Environmental Restoration of Diesel-Range Organics from Project Chariot, Cape Thompson, Alaska - 16147

Cape Thompson, Alaska, is located approximately 120 miles north of Kotzebue, Alaska, within the Ogotoruk Creek valley. The remote Alaskan location is also home to the US Atomic Energy Commission (AEC) Project Chariot study area. Project Chariot was part of the Plowshare Program, created in 1957 by the AEC to study peaceful uses for atomic energy. Project Chariot began in 1958 when a scientific field team chose Cape Thompson as a potential site to excavate a harbor using a series of nuclear explosions. Between 1959 and 1962, AEC was responsible for over 40 pretest bio-environmental and geotechnical studies of the Cape Thompson area. Strong public opposition to the tests developed, which caused the project to be cancelled with no nuclear explosions ever conducted at the site. In the early 1960's, several deep boreholes were drilled to obtain subsurface, temperature, and high-explosive testing data. Refrigerated diesel was used as a drilling fluid to keep the boreholes open so that thermistors could measure subsurface temperatures. The test boreholes were named Able, Baker, Charlie, Dog, and X-1. In the summer of 2014, the US Department of Energy, Office of Legacy Management (LM) completed the remediation of remnant diesel at these sites. Prior to the 2014 remediation activities, LM conducted two separate characterization field studies to determine the nature and extent of the contamination from the 1960's drilling activities. In 2013, LM developed a remediation plan that was submitted and approved by the Alaska Department of Environmental Conservation (ADEC). The 2014 Project Chariot remediation had some unique logistical challenges and historic issues with stakeholders to overcome. Many of the historic issues centered on previous AEC scientific actions and a general mistrust of the government by the local stakeholders. Prior to and during the 2014 remediation work, an outreach campaign was implemented to discuss prior activities at the site and include stakeholders from Point Hope, Alaska, in the remediation-planning process. The remoteness of the region also provided unique logistical challenges. Because of the short field season, LM initiated several subcontracts over a year in advance. The field camp, heavy equipment, and soil storage containers were shipped by barge to the site. Personnel, food, and small materials were flown in by bush plane from Kotzebue, Alaska. Previous site characterization had identified diesel-range organics (DROs) as the primary contaminant. The cleanup levels for the site were provided by the Alaska Administrative Code (AAC) for the Arctic Zone region. The AAC cleanup level for DROs on gravel pads is 500 mg/kg and 12,500 mg/kg tundra in the arctic-zone region. To help determine the extent and direct soil excavation in the field, photoionization detectors (PIDs) were used for field screening. While the PID results provided rough limits for excavation, laboratory confirmation sampling was required to provide actual excavation guidance. All excavation locations remained open until laboratory results met AAC criteria and ADEC approved closure. The laboratory sample protocol required the bush plane to pick up samples almost daily and transfer them to Anchorage for analysis. In order to meet holding time requirements, all laboratory samples were analyzed on an expedited basis. Analytical results and ADEC approval notifications were e-mailed back to the site via satellite connection. A total of 202 laboratory analyses were needed to complete the remediation effort. At each location (Able, Baker, Charlie, Dog, and X-1), the boreholes were plugged and abandoned in accordance with ADEC regulations. Contaminated soil was excavated, and the material was placed in super-sacks and hauled to connex boxes located near the sea shore. Approximately 786 tons of material was removed and shipped by barge to the Columbia Ridge Landfill in Arlington, Oregon, for disposal. The well casings were cut and sealed below ground surface. Each location was graded to its original contours and reclaimed with native vegetation, and a geodetic survey was conducted. Tours with leaders of the Village of Point Hope and the land owner (Arctic Slope Regional Corporation) were conducted prior to the final demobilization. All remediation equipment and camp facilities were removed at the completion of the project. (authors)

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

Corrective Action Decision Document/Corrective Action Plan for Corrective Action Unit 413: Clean Slate II Plutonium Dispersion (TTR) Tonopah Test Range, Nevada. Revision 0

This Corrective Action Decision Document/Corrective Action Plan provides the rationale and supporting information for the selection and implementation of corrective actions at Corrective Action Unit (CAU) 413, Clean Slate II Plutonium Dispersion (TTR). CAU 413 is located on the Tonopah Test Range and includes one corrective action site, TA-23-02CS. CAU 413 consists of the release of radionuclides to the surface and shallow subsurface from the Clean Slate II (CSII) storage–transportation test conducted on May 31, 1963. The CSII test was a non-nuclear detonation of a nuclear device located inside a concrete bunker covered with 2 feet of soil. To facilitate site investigation and the evaluation of data quality objectives decisions, the releases at CAU 413 were divided into seven study groups: 1 Undisturbed Areas 2 Disturbed Areas 3 Sedimentation Areas 4 Former Staging Area 5 Buried Debris 6 Potential Source Material 7 Soil Mounds Corrective action investigation (CAI) activities, as set forth in the CAU 413 Corrective Action Investigation Plan, were performed from June 2015 through May 2016. Radionuclides detected in samples collected during the CAI were used to estimate total effective dose using the Construction Worker exposure scenario. Corrective action was required for areas where total effective dose exceeded, or was assumed to exceed, the radiological final action level (FAL) of 25 millirem per year. The results of the CAI and the assumptions made in the data quality objectives resulted in the following conclusions: The FAL is exceeded in surface soil in SG1, Undisturbed Areas; The FAL is assumed to be exceeded in SG5, Buried Debris, where contaminated debris and soil were buried after the CSII test; The FAL is not exceeded at SG2, SG3, SG4, SG6, or SG7. Because the FAL is exceeded at CAU 413, corrective action is required and corrective action alternatives (CAAs) must be evaluated. For CAU 413, three CAAs were evaluated: no further action, clean closure, and closure in place. The CAAs were evaluated on technical merit focusing on performance, reliability, feasibility, safety, and cost. Based on the evaluation of analytical data from the CAI, review of future and current operations at CAU 413, and the detailed and comparative analysis of CAAs, clean closure was selected as the preferred CAA for CAU 413 by the U.S. Air Force, Nevada Division of Environmental Protection, and U.S. Department of Energy at the CAA meeting held on August 24, 2016.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Phenomena-based Uncertainty Quantification in Predictive Coupled- Physics Reactor Simulations

This project has sought to develop methodologies, tailored to phenomena that govern nuclearreactor behavior, to produce predictions (including uncertainties) for quantities of interest (QOIs) in the simulation of steady-state and transient reactor behavior. Examples of such predictions include, for each QOI, an expected value as well as a distribution around this value and an assessment of how much of the distribution stems from each major source of uncertainty. The project has sought to test its methodologies by comparing against measured experimental outcomes. The main experimental platform has been a 1-MW TRIGA reactor. This is a flexible platform for a wide range of experiments, including steady state with and without temperature feedback, slow transients with and without feedback, and rapid transients with strong feedback. The original plan was for the primary experimental data to come from in-core neutron detectors. We made considerable progress toward this goal but did not get as far along as we had planned. We have designed, developed, installed, and tested vertical guide tubes, each able to accept a detector or stack of detectors that can be moved axially inside the tube, and we have tested several new detector designs. One of these shows considerable promise.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Development of an Improved RELAP5-3D Model for the High Temperature Test Facility

High-temperature gas-cooled reactors (HTGRs) are rapidly approaching deployment. Confidence in transient analysis of these systems requires modeling and simulation tools that have been validated against data relevant to HTGR conditions. The High Temperature Test Facility (HTTF) is an integral effects thermal hydraulics test facility for prismatic HTGRs. In spring and summer of 2019, HTTF was used for a series of experiments that now serve as the basis for the Organization of Economic Cooperation and Development / Nuclear Energy Agency Thermal Hydraulic Code Validation Benchmark for High Temperature Gas-Cooled Reactors using HTTF Data (HTGR T/H Benchmark). Previous analyses as part of the HTGR T/H benchmark used a RELAP5-3D model developed at Idaho National Laboratory (INL) and demonstrated an ability to reproduce trends in the measured data but difficulties reproducing experimental values within their uncertainty. These difficulties were largely attributed to assumptions made during the development of the initial RELAP5-3D model, which predated the HTTF experiments. A significant cause of difficulty reproducing the measured temperatures may be the radial nodalization of the previous RELAP5-3D model. In this paper, we present a new RELAP5-3D model of HTTF with finer radial nodalization built to assess the impact of radial heat transfer. We describe the new model and compare it against the old one at full-power steady state and for the pressurized conduction cooldown (PCC) transient. These analyses are based on the code-to-code comparison exercise for the PCC problem of the HTGR T/H benchmark. We compare maximum block temperature as the primary figure of merit and include discussion on intracore natural circulation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Sister Rod Destructive Examinations (FY23) Appendix F3: Uncertainty and Conservative Bias in the Cyclic Integrated Reversible-Bending Fatigue Test

The development and application of the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) has been documented over the past 10 years. The CIRFT system was developed in collaboration with the U.S. Nuclear Regulatory Commission (NRC) to address the concern that high-burnup fuel rod fatigue performance was degraded compared to cladding-only behavior. Fatigue data measured by CIRFT on high-burnup fuel rods were then used by the NRC to develop a fatigue response curve for pressurized water reactor fuel rods with Zircaloy-4 cladding in NUREG-2224. Subsequently, CIRFT has been used in the sister rod project to test whether reoriented hydrides associated with vacuum drying influenced the fatigue lifetime, to extend the fatigue testing database to include fuel rods with ZIRLO cladding, and to add to the experience on both Zircaloy-4 and M5 cladding. These data will be used as a reference for subsequent characterization of fuel rods removed from the high-burnup demonstration cask to better understand the effects of long-term dry storage, which can be thought of as a multiyear elevated temperature anneal. The objective of fatigue testing is to provide data that can be used to develop a best-estimate fatigue curve and a conservative fatigue design curve that accounts for uncertainty and can be compared with actual transportation conditions. The process for defining a design fatigue curve for commercial used fuel rods was developed by Oak Ridge National Laboratory (ORNL) in FY22. One issue noted in the fatigue evaluation was that the calculated combined uncertainty in the strain amplitude values determined from the measured data was large. It was proposed then to directly measure the uncertainty in the strain amplitude to either validate the calculated combined uncertainty or update the uncertainty based on the measurement data. Thus, the purpose of this appendix is to document the measured uncertainty in the strain amplitude data. In the process, it was determined that there has been a conservative bias in the calculated strain amplitude that is small enough to be within the measurement uncertainty. This observation is discussed herein, and a case is made to continue to include the conservative bias in the data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of Ultra Small Scale Mechanical Testing and Localized He Implantation for Nuclear Applications

Mechanical properties degradation of components deployed in nuclear applications has been a concern since the deployment of first generation nuclear power plants. Radiation damage along with environmental degradation, elevated temperatures, and long term deployment of nuclear structures pose a challenging environment for all materials in these applications. Traditionally, the mechanical properties are assessed by standardized macroscopic testing methods. While those are providing engineering relevant data these techniques do have limitations, especially on reactor irradiated materials, due to the large size of the samples. Limitations on the amount of radioactivity of large samples must be applied for research applications as they are conducted at universities. In addition it can be difficult to retrieve sufficiently large amounts of material from an operating reactor component. Therefore smaller sample sizes are desirable and reduce the amount of radioactive material which needs to be handled or material which needs to be retrieved. In addition, the number of tests which can be performed on materials irradiated in a test reactor program is limited due to limited space in specific dose and temperature positions. Downsizing of the mechanical test can overcome this issue and perform more tests on the same amount of irradiated materials, reducing the statistical uncertainties. In addition, smaller scale samples allow access to specific regions of interest in a material or component, be it the zone in front of a crack tip or the sampling of very small components like springs or welded areas. In addition ion beam irradiation has become an integral part of nuclear materials research to enhance the understanding of radiation damage in materials in a time and cost effective fashion. The drawback of ion beam irradiation is the small penetration depth of the ions into the material making localized microstructural characterization as well as small scale mechanical testing necessary. Another more fundamental motivation is the fact that since the beginning of electron microscopy, one has been concerned about imaging and quantifying defects in materials. Multiscale tools to image a large range of radiation induced defects have been established starting from atom probe tomography to transmission electron microscopy, up to scattering techniques. Mechanical testing, however, has always been performed on a large scale, and an attempt to match the same scale where a defect is observed with the materials' mechanical property response has not yet been realized. However, the question of separate effects testing on specific radiation induced defects remains and is an additional driving force for small scale mechanical testing. In this work we present an approach for small scale mechanical testing in irradiated materials and the variety of possible test methods developed today. We also present recent efforts of performing localized ion beam implantation to achieve a number of different doses in the same region of interest enhancing comparability and quantification of observed phenomena. It is the intention to give the reader an overview of the different microscale mechanical testing techniques possible today as well as an introduction of a novel ion beam irradiation method using a newly established capability. In this paper we can't discuss all aspects of all small scale mechanical testing of irradiated materials However, It does give an overview over the techniques enabled due to recent effort and enhanced instrument capability. While indentation or micro compression testing has been performed previously tensile testing and ion beam irradiation using the He ion beam microscope are novel approaches not attempted previously. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2023

Film condensation with high heat fluxes and scaled experiments using pure steam for reactor containment cooling

Condensation tests were performed using a newly developed test facility for scaling the passive containment cooling system (PCCS) to a small modular reactor (SMR). The PCCS of the SMR plays a pivotal role in ensuring greater safety, reliability, and compactness than what is afforded by traditional reactors. Therefore, a well-designed PCCS is essential to SMRs. However, previous studies and test data were unsuitable for scaling, due to high variation in the test geometry and operating conditions. This study intends to close this research gap by using a novel designed scaled test facility consisting of vertical condensing test sections featuring 1-, 2-, and 4-inch-diameter condensing tubes with annular water cooling, and by applying superheated and saturated steam with different steam mass flow ranges of 5–25 g/s. Further, the primary test data, including axial temperatures, mass flow rates, and pressures, were used in conjunction with a standard data reduction method to estimate critical parameters such as heat fluxes, heat transfer coefficients, and condensation rates. These scaled test data would support improving empirical correlations and validating condensation models to identify scaling distortion for SMR PCCSs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Application of Dynamical System Scaling for Accelerated Fuel Qualification Efforts

There are considerable ongoing research and testing campaigns to qualify new fuel designs such as metallic fuels for advanced reactor designs and the Accident Tolerant Fuel (ATF) campaign for light water reactors (LWR). The typical research and development lifecycle needed to qualify a nuclear fuel design, under ideal conditions, can take up to 20-25 years which limits the ability for new fuels to make fast deployments into commercial, test, and research reactors. While there are several efforts to accelerate nuclear fuel qualification through advanced modeling coupled with state-of-the-art experiments, there is not yet a framework to methodically quantify and rank distortions occurring in experimental test specimens needed to validate nuclear fuel performance codes. This study uses the Dynamical System Scaling (DSS) methodology to quantify transient scaling distortions that occur over experiment and simulated transients. DSS analysis metrics provide a researcher with several tools and information embedded within test data to identify dominant phenomena, associated timescales, and quantify a transient?s overall dynamic distortion. This is demonstrated with separate analyses of the SETH-C and CHF-SERTTA thermal-hydraulic experiments in the TREAT facility at Idaho National Laboratory. The outcome of this work is a scaling and data analysis approach to account for the influence a scaled fuel test specimen?s geometric and temporal distortions have on its ability to be representative of the full-scale design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

AGC-4 Specimen Post-Irradiation Examination Data Interim Report

This interim report documents preliminary results of the post-irradiation examination material property testing from the fourth advanced graphite creep (AGC), AGC 4, capsule specimens. This is the fourth of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades to moderate dose levels (=7 dpa). The AGC 4 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 800°C and to a peak dose of 8 dpa. Half of the AGC-4 specimens were subjected to compressive stresses to induce irradiation creep. Post-irradiation testing and measurement results are reported with the exception of thermal testing, which is still in progress, and irradiation mechanical strength testing. Additionally, some specimens initially deemed too hot to be examined in the ART Graphite laboratory may still be measured. The data reported includes specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and dimensional data necessary to derive density, elastic constants (Young?s modulus, shear modulus, and Poisson?s ratio) from ultrasonic time of flight velocity measurements, Young?s modulus from the fundamental frequency of vibration, and electrical resistivity. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to the irradiation environment and applied load on the specimens, will be discussed later in AGC 4 post-irradiation examination analysis reports.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Key results from examinations of seven high burnup pressurized water reactor spent nuclear fuel rods

At present, spent nuclear fuel (SNF) assemblies discharged from US commercial power plants are placed into dry storage following a short cooling time (<10 years) in the plant’s spent fuel pool. The process of packaging the spent fuel into dry-storage canisters includes a drying step to remove residual water from the canister. During the drying process, the fuel rod cladding may reach temperatures as high as 400°C. Oak Ridge National Laboratory (ORNL) is performing destructive examinations of high burnup (HBU) (>45 GWd/MTU) SNF rods to address knowledge and data gaps related to extended interim storage and eventual transportation for disposal. The rods examined include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Three rods were subjected to a thermal transient to assess the effects of decay-heat-driven high temperatures expected during vacuum drying of the fuel as it is prepared for interim dry storage. The examinations focus on the composite fuel rod performance, as compared with the performance of defueled rod cladding, and establish the baseline mechanical properties of a fuel rod before interim dry storage. The key results of these examinations are presented, including the measured mechanical and fatigue properties, observations of cladding hydrogen pickup and hydride reorientation effects on rod performance, effects of the simulated drying temperatures on rod performance, and general conclusions of SNF performance in extended interim dry storage and transport. The rods were found to be strong and durable in the expected loading conditions, even considering the formation of radial hydrides associated with vacuum drying. The combined testing provides a broad body of data supporting extended interim storage and transportation performance of HBU spent fuel.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Recent SAM Code Improvement to Heat Transfer Modeling Capabilities

System Analysis Module (SAM) is currently under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor (non-LWR) safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report summarizes recent SAM code improvement to heat transfer modeling capabilities to support system and core thermal-hydraulics analysis for advanced nuclear reactor systems. These new capabilities include implementation of the energy equation for solid structure in porous medium and its coupling with porous medium flow energy equation; coupling between one-dimensional flow equation and three-dimensional heat conduction equation; coupling between two-dimensional heat structure and three-dimensional porous medium flow equations; and implementation of effective thermal conductivity correlations for solid pebble beds and its validation against High Temperature Test Unit (HTTU) experimental data. Demonstration or validation test problems are also included in this report to confirm the successful implementation of these capabilities.

1D/3D coupling↗
DOE OSTI2017

Baseline Characterization Database Verification Report – PCEA Billet 02S8 7

The purpose of this Engineering Calculations and Analysis Report is to present the data being collected in the Baseline Graphite Characterization program, which is directly tasked with supporting the Idaho National Laboratory’s (INL’s) research and development efforts on the Advanced Reactor Technologies (ART) Program. This program is populating a comprehensive database that will reflect the baseline properties of nuclear grade graphite with regard to individual grade, billet, and position within individual billets. The physical and mechanical property information being collected will be transferred to the Nuclear Data Management and Analysis System (NDMAS), and that database will help populate the handbook of property data available to member nations of the Generation IV International Forum. Transfer of these data from the applicable technical lead to the dissemination databases available to other end users requires a full review of the test procedures and data collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents the analysis for PCEA billet 02S8 7 and facilitates release of associated data to the NDMAS custodians.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Interlaboratory round robin study on axial tensile properties of SiC-SiC CMC tubular test specimens [Interlaboratory round robin study on axial tensile properties of SiC/SiC tubes]

An interlaboratory round robin study was conducted on the tensile strength of SiC–SiC ceramic matrix composite (CMC) tubular test specimens at room temperature with the objective of expanding the database of mechanical properties of nuclear grade SiC–SiC and establishing the precision and bias statement for standard test method ASTM C1773. The mechanical properties statistics from the round robin study and the precision statistics and precision statement are presented herein. The data show reasonable consistency across the laboratories, indicating that the current C1773–13 ASTM standard is adequate for testing ceramic fiber reinforced ceramic matrix composite tubular test specimen. Furthermore, it was found that the distribution of ultimate tensile strength data was best described with a two–parameter Weibull distribution, while a lognormal distribution provided a good description of the distribution of proportional limit stress data.

36 MATERIALS SCIENCE↗
DOE OSTI2023

An Initial Evaluation of the Elevated-Temperature Cyclic Properties of Optimized 316H Stainless Steel Fabricated by Powder Metallurgy Hot Isostatic Pressing

Powder metallurgy hot isostatic pressing (PM-HIP) is a mature advanced manufacturing technology, which can consolidate metallic powder under high temperature and pressure to generate near-net shape components. This process can lower production costs and reduce component lead times for nuclear reactor construction and can be readily deployed in the near term for microreactor components because of their smaller size. The purpose of this work is to continue the evaluation of elevated-temperature cyclic material properties for PM-HIP 316H stainless steel. Prior scoping results have shown that the commercially procured PM-HIP stainless-steel materials have reduced creep-fatigue performance compared to wrought 316 stainless steels. To better understand how PM-HIP processed material properties compare to traditional manufacturing methods, additional billets of PM-HIP 316H stainless steel were procured and tested. This testing is necessary to understand how elevated-temperature properties are influenced by the PM-HIP process and what data are needed for incorporating PM HIP 316H into the American Society of Mechanical Engineers Boiler Pressure Vessel Code, Section III, Division 5 for high temperature reactor construction. Specifically, an additional PM-HIP 316H stainless steel billet with lower oxygen and nitrogen concentrations was fabricated by the Nuclear Advanced Manufacturing Research Centre of the United Kingdom and tested to understand if elevated-temperature cyclic properties can be improved. The creep-fatigue resistance of the material from this new billet did not show appreciable improvement compared to prior testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗