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At least 1,207 records · Page 67

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NASA NTRS1974

Graphite thermal expansion reference for high temperature

The design requirements of the aerospace and high-temperature nuclear reactor industries necessitate reliable thermal expansion data for graphite and other carbonaceous materials. The feasibility of an acceptable reference for calibration of expansion measuring systems that operate in carbon-rich atmospheres at temperatures ranging to 2500 C is the prime subject of this work. Present-day graphite technology provides acceptable materials for stable, reproducible references, as reflected by some of the candidate materials. The repeatability for a single specimen in a given expansion measuring system was found to be plus or minus 1%, while the combined results of several tests made on a number of samples fell within a plus or minus 2.5% band.

Gaal, P. S.↗

DOE OSTI2023

Fact Sheet - Long-length Scintillating Fibers for Radiation Detection

As the world looks to underground geological repositories for storing nuclear waste, technologies to safeguard the material are needed. An example of a project that developed a new technology is Tripwire. Tripwire proposes a multi-sensor system approach for geological spent nuclear fuel repositories that relies of radiation, vibration, and electromagnetic detection. Tripwire was developed by the Applied Radiation Measurements and Systems (ARMS) group at INL and sponsored by the National Nuclear Security Administration (NNSA). The ARMS group performs research and development, testing and evaluation, operational support, and training focused on applied ionizing radiation detection and measurement. Application areas include nuclear engineering for advanced reactors and fuel cycle operations, nuclear nonproliferation, nuclear counterproliferation, nuclear forensics, and arms control and disarmament.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

DOE OSTI2020

Single-phase catalysis for reductive etherification of diesel bioblendstocks

We developed a single-phase Pd/NbOPO 4 catalyst for reductive etherification that displays high catalytic activity, product selectivity, and regeneration stability.

Hafenstine, Glenn R.↗

DOE OSTI2021

Catalytic Reduction of Residual Oxygen for the Purification of Carbon Dioxide Streams from Pressurized Oxy-Combustors

Pressurized oxy-combustion (OC) is regarded as an energy-efficient alternative to atmospheric OC. To utilize the captured carbon dioxide (CO2) from pressurized OC for enhanced oil recovery or carbon storage in geological formations, the CO2 product stream must meet the required purity specifications. One major impurity in OC flue gases is residual O2 (e.g., 3 vol%). In this study, a catalytic approach has been investigated to remove the residual O2 through a reduction reaction with injected methane (CH4) in a reverse-flow fixed-bed (RFFB) reactor. The performance was tested for a supported copper (Cu)-based catalyst synthesized by incipient wetness impregnation and a cobalt-manganese (CoMn) oxide catalyst synthesized by a coprecipitation method. A commercial Cu-based catalyst was also studied as the reference. Experimental results revealed that under the reverse-flow operation, the temperature profiles in the reactor were symmetrical on both sides of the catalyst zone and were more homogenous than those under the conventional one-direction-flow operation. Among the tested catalysts, the CoMn catalyst was the most reactive for promoting the reduction of O2 with CH4, able to attain >95% O2 removal at 475 °C and a gas hourly space velocity (GHSV) of 18,000 h-1 with a feed gas containing 3 vol% O2. Under all experiments, the concentrations of carbon monoxide (CO) were lower than 30 ppmv in the effluent gas streams. The reaction between O2 and CH4 could be depicted by using the Mars–van Krevelen redox mechanism based on a two-step reduction–oxidation reaction pathway.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Decoding Early Candidacy of High Entropy Alloys for Nuclear Application using the Advanced Test Reactor through Predictive Methods and Combinatorial Testing

High Entropy Alloys (HEAs) are identified candidates for nuclear applications owing to their superb mechanical and thermal properties. In line with evaluating their candidacy, a challenge remains in their validity as structural replacements for extreme environments. Each compositional graded specimen consists up to four different compositions spanning the fueled zone for non-prototypical neutron irradiation testing in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). We have fabricated a set of four compositionally graded specimens using spark plasma sintering that include at least five of the following elements: Cr, C, Al, Zr, Mo, Nb, Ta, V, Ti, W, and Fe. In time for this meeting the role of temperature expected to play a role in-pile cladding chemical interactions (FCCI), mechanical interactions (FCMI), irradiation damage, creep, and resistance will be reported. The irradiated portion has a scheduled irradiation date in April 2020 and with post irradiation examination (PIE) to follow.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ~400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α'), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Prospects for additive manufacturing of nuclear fuel forms

Additive manufacturing (AM) is viewed as a critical enabling technology for achieving superior performance and improved economics across many sectors. Although the past decade has seen increased application of AM methods to nuclear reactor cladding and structural materials, exploration of AM as applied to the uranium-bearing nuclear fuel forms has been limited. The two major families of nuclear fuel forms (monolithic, particle/dispersion) are utilized with their own set of core objectives in mind, and their differences require distinct fabrication infrastructures. These reference fabrication methods impose many limitations on nuclear fuels. Further, both currently operating reactors and future concepts have the potential for improved performance if these accepted limitations are relaxed or removed entirely. The primary limitations of reference fabrication processes for the common nuclear fuel forms are outlined in this paper. This groundwork is then used to identify avenues of fuel performance specific to each of these fuel architectures that could be exploited if the restrictions of conventional fuel fabrication are removed. Moreover, multiple targets for AM studies are laid out for each of the major nuclear fuel variants. Finally, key strategic components to guide research activities in AM of nuclear fuels are outlined, with an emphasis on use of modeling and simulation to motivate research aims and embrace of an accelerated testing methodology to screen and quality new fuel forms.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Superconducting Magnet Qualification Methodology for the Material Plasma Exposure Experiment

The Material Plasma Exposure eXperiment (MPEX), currently under design, is a new linear plasma device to advance the understanding of plasma-material interactions through the generation and delivery of plasmas as they are expected in future fusion reactor divertors. MPEX will be a steady-state device to study high-fluence exposures of plasma-facing materials and components. The requirements for the magnetic field at the target and the heating stages make the application of superconducting coils necessary. Conceptual designs for the superconducting magnets have been developed, and multiple cryostats with warm bore diameters of either 65 cm or 156 cm are envisioned to facilitate their integrated and timely assembly with other systems such as vacuum, water cooling, and RF power. Although design, fabrication, and testing for the magnets as stand-alone units are straightforward, challenges will arise during the integration of the system. Two different field profiles will be used during operation. The magnetic field where the electron cyclotron heating occurs needs to operate at both 1.25 and 2.5 T. It is critical that the magnets all share the same magnetic axis and alignment. The mutual inductance between cryostats will affect the quench behavior of the system. Also, cryostat-to-cryostat forces can be as large as 700 kN, and the magnitude and direction will change depending on which coils are energized. The design of the system must take those characteristics into account along with the quench scenarios. This paper describes the qualification approach that will be used to determine whether stand-alone tests can be used to ensure the success of the integrated system. Fiducials will be used to define the location of the magnetic axis for each cryostat to ensure proper alignment. Quench tests of a single magnet will be performed at a current above the normal operating current to account for additional stored energy from the mutual inductance to adjacent cryostats. Also, a 1018 steel plate will be mounted on either end of a cryostat to simulate the cryostat-to-cryostat forces. Requirements for the size and location of the steel plates are described.

Burkhardt, Earle↗

NASA NTRS2007

Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a representative lunar surface reactor shield design is evaluated at various power levels in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to anchor a CFD model. Performance of a water shield on the lunar surface is then predicted by CFD models anchored to test data. The accompanying viewgraph presentation includes the following topics: 1) Testbed Configuration; 2) Core Heater Placement and Instrumentation; 3) Thermocouple Placement; 4) Core Thermocouple Placement; 5) Outer Tank Thermocouple Placement; 6) Integrated Testbed; 7) Methodology; 8) Experimental Results: Core Temperatures; 9) Experimental Results; Outer Tank Temperatures; 10) CFD Modeling; 11) CFD Model: Anchored to Experimental Results (1-g); 12) CFD MOdel: Prediction for 1/6-g; and 13) CFD Model: Comparison of 1-g to 1/6-g.

Pearson, J. Boise↗

DOE OSTI2019

Advanced Fuels Campaign 2019 Accomplishments

The mission of the Advanced Fuels Campaign (AFC) is to perform or support research, development, and demonstration (RD&D) activities to identify and mature innovative fuels and cladding materials and associated technologies with the potential to improve the performance and safety of current and future reactors; increase the efficient utilization of nuclear energy resources; contribute to enhancing proliferation resistance of the nuclear fuel cycle; and address challenges related to waste management and ultimate disposal. AFC pursues its mission objectives using a goal-oriented, science-based approach that seeks to establish a fundamental understanding of fuel and cladding behaviors under conditions that arise during fabrication, normal steady-state irradiation, off-normal transient scenarios, and storage/disposal. This approach includes advancing the theoretical understanding of fuel behavior, conducting fundamental and integral experiments, and supporting the mechanistic, multi-scale modeling of nuclear fuels to inform and guide fuel development projects, advance the technological readiness of promising fuel candidates, and ultimately support fuel qualification and licensing initiatives. In the area of advanced tools for the modeling and simulation of nuclear fuels, the AFC works in close partnership with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, participating with NEAMS in developing mechanistic fuel behavior models and providing experimental data to inform and validate its most advanced tools. Specifically, AFC objectives in thecoming five year horizon include: 1. support the industry-led development of Accident Tolerant Fuel (ATF) technologies with improved reliability and performance under normal operations and enhanced tolerance during hypothetical accident scenarios, with implementation of lead test rods/assemblies of one or more ATF concepts in commercial reactor(s) by or before 2022 and batch reloads of near-term concept(s) in the 2023-2026 timeframe; 2. lead research and development on innovative fuel and cladding technologies for applications to future advanced reactors, especially fast-spectrum reactors, including reactors that utilize both once-through and recycle scenarios; 3. continue the development and demonstration of a multi-scale,science-based approach to fuel development and testing, andcontribute to the establishment of a state-of-the-art R&D infrastructure necessary to accelerate the development of new fuel concepts; and collaborate with NEAMS on the development and validation of multi-scale, multi-physics, and increasingly predictive fuel performance models and codes. 4. collaborate with NEAMS on the development and validation of multi-scale, multi-physics, and increasingly predictive fuel performance models and codes.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Effects of Internal and External Heat Sources on Cladding Microstructure and Rupture Performance

In the event of a Loss of Coolant Accident (LOCA), the primary supply of cooling water for a nuclear reactor is lost, leading to a significant pressure differential across the cladding wall. Without adequate cooling, the fuel rods continue to heat as a result of fission reactions. Research at Oak Ridge National Laboratory’s (ORNL’s) Severe Accident Test Station (SATS) is currently focused on evaluating fuel cladding performance using an external infrared lamp as a heat source, whereas legacy testing primarily utilized an internal heating approach. While external heating may better simulate the effect of neighboring fuel rods heating a central rod, it may not accurately represent the internal heat absorption from the fuel during accident transients. The heating dynamics depend greatly on the fuel rod's position within the bundle and the reactor. To fully understand the implications of a LOCA event and assess the influence of internal heating on cladding performance, combined internal heating and pressurization capability was developed at ORNL. Tests were conducted to compare cladding segments heated internally (representing heat from the fuel within the rod) and externally (representing heat from adjacent fuel rods). The findings indicate that both internal and external heating result in comparable rupture temperatures during 5°C/s laboratory LOCA tests and also agree with legacy test data. Axial temperature gradients and internal heat source dispersal were found to significantly impact cladding deformation and rupture geometry. Clear modifications were outlined to further improve the capability with heating rates above 5°C/s.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2010

Field Testing of a Pneumatic Regolith Feed System During a 2010 ISRU Field Campaign on Mauna Kea, Hawaii

Lunar In Situ Resource Utilization (ISRU) consists of a number of tasks starting with mining of lunar regolith, followed by the transfer of regolith to an oxygen extraction reactor and finally processing the regolith and storing of extracted oxygen. The transfer of regolith from the regolith hopper at the ground level to an oxygen extraction reactor many feet above the surface could be accomplished in different ways, including using a mechanical auger, bucket ladder system or a pneumatic system. The latter system is commonly used on earth when moving granular materials since it offers high reliability and simplicity of operation. In this paper, we describe a pneumatic regolith feed system, delivering feedstock to a Carbothermal reactor and lessons learned from deploying the system during the 2010 ISRU field campaign on the Mauna Kea, Hawaii.

Craft, Jack↗

DOE OSTI2020

Design of a Critical Experiment to Validate Yttrium Hydride at Varying Temperatures

As microreactors become an emerging nuclear energy source for remote applications, designers are looking at the use of high temperature moderator material to reduce the fuel mass in the core by up to a factor of two. The microreactor community is in constant search of the best moderator material to retain hydrogen at high temperatures while optimizing neutronic performance. Recent simulation work has shown that materials which have a high and low Z component are optimal for this application. Material properties studies have further shown that yttrium hydride maintains hydrogen and stays as a solid until well over the needed temperatures, and thus retains hydrogen. A critical experiment is planned at the National Criticality Experiments Research Center (NCERC) to test the reactivity behavior of yttrium hydride at standard and elevated temperatures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Expansion of the Fuel Motion Monitoring System

The Idaho National Laboratory (INL) Transient Reactor Test Facility (TREAT) Fuel Motion Monitoring System (FMMS) is a diagnostic tool used to visualize the movement and location of fuel within the nuclear reactor during transient experiments. TREAT is designed to test the behavior of nuclear fuel under accident conditions, such as a rapid power increase. The TREAT FMMS, also referred to as the “hodoscope” because of the system’s massive steel collimator component, is located on the North beam port of the reactor. While the FMMS system is capable of incorporating a total of 360 sensors, initial refurbishment efforts resulted in a array of 96 fast neutron detectors. This initial array was configured to provide a narrow (2 channels) full-length view of experiment vessels with an extended 4-channel–wide region in the center of the array capable of providing a full view of all initially planned experiments. Planning and efforts to expand the FMMS array began immediately following the restart of TREAT. An additional 96 detectors were evaluated, characterized, and prepared for installation at TREAT during the summer of 2018. Unfortunately, funding reductions forced the project towards a 5-year hiatus. In October of 2023 funding from the Department of Energy’s Nuclear Science User Facilities enabled the expansion to recommence. This report serves to document the activities performed to expand the FMMS detector array to 192 channels in order to provide a broader view of the larger and more sophisticated test capsules currently planned for future irradiation in the TREAT reactor. This expansion also included doubling the data acquisition capability, addressing the new heat load on the system, and synchronizing the time for all digital components.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

TRISO Fuel Development and Qualification in the US

Presentation of TRISO Fuel Development and Qualification in the US. Including Status of HTGR and TRISO Technology Circa 2000, Advanced Gas Reactor focus on licensing and domestic TRISO fuel fabrication capabilities. Irradiation experiments discussed include AGR-1, AGR-2, AGR-5/6/7, and AGR-3/4. Results, of tests and Kernel and coating behaviors during irradiation, Fission product release from UCO Fuel Compacts, study of failed particles, safety and accident performance summaries, and Parfume modeling and capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2006

Preliminary Analysis of a Water Shield for a Surface Power Reactor

Pearson, J. Boise↗

DOE OSTI2022

Comparative Structural Rigidity Analysis of the NBSR DDE and the NBSR LEU Fuel Element

The National Bureau of Standards Reactor (NBSR) is a 20-MW heavy-water moderated and cooled, enriched-fuel, tank-type reactor located in the NIST Center for Neutron Research laboratory complex. NBSR is one of six U.S. high performance research reactors (USHPRR), including one critical facility, that actively collaborates with the National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program to convert from the use of highly enriched uranium fuel to the use of low-enriched uranium (LEU, < 20 wt% U-235) fuel. On the LEU U-10Mo fuel qualification side, mini-plate and large-plate irradiations have been successfully performed, and more experiments are either ongoing or planned. As an additional experimental campaign to support licensing of the LEU fuel for the use in the NBSR, the NBSR Design Demonstration Element (DDE) will be a full-sized test element similar to the NBSR LEU fuel element and irradiated in a test reactor under conditions as prototypic of the NBSR LEU fuel element operating conditions as possible.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Second group of irradiation capsules: property data of irradiated welded stainless steel 347 and irradiated, welded, and hydrogen charged Zircaloy-4 for SHINE

Zircaloy-4 and AISI 347 stainless steel were irradiated to low doses (10 20 and 10 21 n/cm 2 ) at low temperature (~60°C) in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). These materials were under consideration for some of the structural components of the medical isotope production facility developed by SHINE Technologies. Zircaloy-4 base metal, as welded, welded and post-weld heat treated (PWHT), and hydrogen charged samples were irradiated. AISI 347 base metal from Sandmeyer Steel, Penn Stainless, and Rolled Alloys suppliers were irradiated. AISI 347 gas tungsten arc welding (GTAW) and flux-cored arc welding (FCAW) samples of Penn Stainless and Rolled Alloys were also irradiated. The irradiated materials were tested with microhardness and tensile tests at room temperature in air. The AISI 347 materials had more elongation after irradiation than the Zircaloy-4 materials. In fact, the Zircaloy-4 welded without PWHT and with H charging had the lowest elongation of any of the materials. This report is the final in a series of reports on this project.

36 MATERIALS SCIENCE↗