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At least 649 records · Page 36
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Embedding Sensors in 3D Printed Metal Structures

The Transformational Challenge Reactor (TCR) program is leveraging recent advances in modeling and simulation, materials, and additive manufacturing (AM) technologies to design a modern nuclear reactor. Some of the main TCR technologies include in situ monitoring and the integration of sensors during the manufacturing of quality-significant nuclear reactor components. This report describes the general procedure and process optimization for embedding sensors within generic stainless steel 316 (SS316) components using laser powder bed fusion (LPBF). A more detailed, quality-significant test plan and supporting procedures are available upon request (ORNL/TM-2021/2127). LPBF involves the use of a scanning laser to selectively melt regions of a powder bed, additively building a part layer by layer. This report describes the LPBF processing technique and discusses the effects of LPBF processing parameters on the success of the sensor embedding process. Experiments used machined cavities in the form of channels in an SS316 base for the sensors to lay in while material is additively built over the top, thereby embedding them in an SS316 matrix. A preliminary investigation involved using empty SS316 sheaths as surrogates to explore the effects of various LPBF processing parameters and the dimensional requirements of the machined channels. Microstructural investigations showed that a smaller channel width/depth combination closer to the sensor’s diameter was best for the embedding process. After the desired parameters were selected, Type-K thermocouples were embedded and evaluated post-embedding using nondestructive thermal testing, as well as destructive sectioning and microscopy. Post-build characterization showed that the thermocouples were well-bonded to the SS316 matrix and were fully functional after embedding. During thermal testing to temperatures up to 500 °C, the embedded thermocouples read consistently with one another and deviated only slightly from the readings of a nonembedded thermocouple located within the furnace. This slight discrepancy was most likely due to differences in the thermal time constants for a nonembedded thermocouple vs. a thermocouple embedded in a solid SS316 block. The results presented in this report will serve as the foundation for future work that will focus on embedding sensors in relevant TCR reactor components and eventually testing those components under neutron irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Neutronic Consideration of TREAT Facility Fuel SiC Recladding

The Transient Reactor Test (TREAT) Facility is an air-cooled reactor composed of graphite-urania fuel encapsulated in zirconium alloy canisters and surrounded by ~2 ft (~61 cm) of nuclear grade graphite. The primary role of TREAT is to provide in-pile nuclear-heated safety research via transient testing [1]. Benchmark efforts have demonstrated sensitivity of TREAT neutronics models to graphite cross section data and key neutron poisons found in the fuel assemblies [2]. During fabrication of TREAT fuel assemblies, a small concentration of boron was inadvertently introduced with variable concentration into the aggregate of fuel blocks prepared for TREAT use. The concentration of boron, and its respective uncertainty, represent the most sensitive component in a neutronics study; unfortunately, nothing can be done to reduce or otherwise mitigate its impact. Hydrogen in the fuel blocks reduces the mean free path of the core and its true quantity and uncertainty is approximated. The hydrogen remains from pitch binder used in the fuel fabrication process and was not completely released due to incomplete fuel graphitization. Destructive fuel testing is necessary to quantify the actual hydrogen content and to improve benchmark model quality. A third neutron poison of some significance is the hafnium impurity in the Zircaloy-3 (Zry-3) assembly cladding. The quantity of hafnium is expected to be within nominal manufacturing limits for nuclear grade zircaloy. Recladding the fuel assemblies with material that is less neutron absorbing is expected to benefit reactor neutron economy. Moreover, TREAT’s total transient energy is restricted based on the rate of Zry-3 oxidation in TREAT’s air coolant environment. Hence, cladding with more benign air reaction kinetics could dramatically uprate the facility’s experimental capability. Silicon carbide (SiC) has been proposed as a candidate material for recladding TREAT assemblies, as SiC/SiC demonstrate favorable properties within nuclear environments [3]. The current operational limits of TREAT are restricted by the oxidation of Zry-3. Change of the cladding to SiC could feasibly double the current temperature limitation. Recladding opportunities allow for previously damaged fuel assemblies to be returned to service, as Zry-3 is not commercially available. The initial effort required to consider proceeding with any recladding effort is to neutronically characterize key reactor parameters while replacing Zry-3 with SiC. A characterization summary, with respective results, is provided herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

MEITNER Resource Team Modeling and Simulation Support to Holos-Quad Reactor Development (Final CRADA Report)

The main objective of this CRADA is to provide Argonne National Laboratory (ANL)’s modeling and simulation capabilities via ARPA-E's MEITNER (Modeling-Enhanced Innovations Trailblazing Nuclear Energy Reinvigoration) Resource Team (RT) arrangement to support the demonstration of the viability of HolosGen’s Holos-Quad reactor design. The Holos-Quad reactor design is an advanced reactor concept that incorporates many new design features, such as the neutron-coupled Subcritical Power Modules (SPMs) in its core design, elimination of balance of plant (BOP) by direct integration of a helium Brayton cycle power conversion system with each SPM, among many other innovative features. The demonstration of the viability of such an innovative reactor design warrants iterations of modeling and simulation and testing. The purpose of this project is to utilize ANL’s modeling and simulation capabilities in nuclear reactor analysis and power conversion system analysis to inform HolosGen and the Design Team (DT) in the design and optimization of the Holos-Quad concept. The major work scopes of this project include: to investigate conceptual designs and materials for radiation shielding to protect personnel and internal components such as turbomachinery; Assess the the helium Brayton cycle power conversion system performance in both nominal and load following conditions; Investigate power conversion components and overall system performance; Identify control strategies to enable load following; Perform simulations of HolosGen’s subscale simulator and using available test data for code validation/benchmark purpose; Perform core thermal-hydraulics, safety analysis, and structural analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Mechanical Properties and Deformation Behavior of Additively Manufactured 316L Stainless Steel (FY2020)

The Transformational Challenge Reactor (TCR) program plans to build most of the TCR core components through additive manufacturing (AM) processes. These processes include laser powder bed fusion (LPBF) for the metallic (316L) components and the newly developed combined process of binderjet printing and chemical vapor infiltration (CVI) for the SiC fuel matrix. Mechanical testing and characterization tasks have been carried out since the beginning of the TCR program to (1) build a property database for the AM materials that will be used in TCR core and (2) to assess the materials’ performance in TCR-relevant conditions. This document reports the outcome of the testing and characterization efforts for the fiscal year with a focus on the mechanical performance data of AM 316L stainless steel (SS). Baseline tensile testing over a wide temperature range of room temperature–600 °C was completed for the AM 316L alloy in as-built, stress-relieved, and solution-annealed conditions. The as-built 316L showed the highest strength, and the alloy after the post-build treatments showed reduced strengths in the low-strain range. However, the strength differences among the AM materials became insignificant in the later part of deformation. Furthermore, regardless of post-build processing, the AM 316L SS showed higher strength and comparable ductility when compared with wrought 316L SS. Thermal creep testing and microstructural evolution during creep deformation were also performed under selected conditions. It was found that the AM 316L steel showed the best creep resistance in the stress-relieved condition. In-situ tensile tests were performed using scanning electron microscopy and 1-ID beamline at the Advanced Photon Source to elucidate the deformation and fracture behavior of AM 316L and the evolution of crystalline stress, dislocations, and pore distribution. Using these in-situ testing data, an in-depth analysis of the roles of microstructural features in deformation and fracture processes is presented herein. The final section of the document introduces ongoing and future activities for materials testing and characterization, including irradiation effect studies and ball punch testing on AM 316L and AM IN718 alloys.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Example Evaluation of a Representative Heat Pipe Test Article Design for Structural Acceptability using ASME Design Rules

This report describes a design analysis of a test article mimicking a core block of a heat-pipe mi-croreactor using the ASME Boiler & Pressure Vessel Code Section III, Division 5, Subsection HB, Subpart B rules. These rules cover the design and construction of elevated temperature nu-clear reactor structural components, like the core block of a heat-pipe microreactor emulated by the test article. The purposes of the report are to: (1) provide a step-by-step example of applying the ASME rules to a reasonable core block geometry and corresponding loading conditions, (2) to provide design feedback on the test article, and (3) to evaluate the ASME rules for use in the design of complex heat-pipe microreactor components. The test article design has a relatively short ASME design life, limited by high thermal stresses. The current ASME rules are adequate for evaluating the core block component designs, though they could be optimized to increase the design efficiency and reduce the amount of designer/analyst time required to execute the design process. The report covers both the current, 2019 edition ASME rules and modified, currently unapproved, rules designed to improve the design process for components with complex geometries and high secondary stresses, like the core block test article.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Corrosion Testing Of Additively Manufactured Stainless Steel 316H In Molten Salt Environments

The development of a new ASTM standard for evaluating the corrosion resistance of additive manufactured (AM) stainless steel (SS) 316H in chloride molten salts is critical for the use of these materials in extreme environments such as molten salt reactors (MSRs). This report pertains to a work package of the Advaned Materials and Manufacturing Technologies (AMMT) developing a systematic methodology to link changes in AM fabrication parameters, namely surface finishing, porosity, microstructure, and chemical heterogeneity, to corrosion performance in NaCl-MgCl2 salt, a proposed secondary coolant for MSRs. During Fiscal Year 2024, Idaho National Laboratory investigators in the AMMT program, utilized SS316H bars, fabricated with laser bed powder fusion at Los Alamos National Laboratory, to establish and optimize the workflow for evaluating these process-to-performance relationships. Standard practices for specimen preparation were established using these specimens, with particular focus on descaling and sectioning methods that align with ASTM guidelines. A comprehensive experimental design for static corrosion testing was developed, with pre- and post-exposure analysis utilizing optical microscopy and scanning electron microscopy. So far standard descaling techniques were optimized, one static corrosion test was conducted on the LANL AM SS316H specimens, and pre- and post-corrosion practices were established. In addition, the work package yielded a review paper on corrosion testing gaps for AM materials in nuclear applications and submitted a proposal for a rapid-turnaround experiment to the Nuclear Science User Facilities Program to investigate the combined effects of proton irradiation and corrosion on AM SS316H. This work package establishes a foundation for evaluating processing-to-performance relationships for AM SS316H in harsh conditions, contributing to the safe and efficient design of components for next-generation nuclear reactors. The creation of a standardized methodology and the generation of relevant publications and future research pathways represent significant strides towards integrating AM materials into critical applications where corrosion resistance is paramount.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Research and Test Reactor Fuels

PRO-RR is the research reactor focused program element of the broader Proliferation Resistance Optimization program (PRO-X) under the National Nuclear Safety Administration (NNSA) in the U.S. Department of Energy (DOE). PRO-X provides a framework for integrating proliferation resistance in nuclear system designs to minimize weapons usable nuclear materials (WUNM) production and diversion pathways while optimizing systems performance for peaceful use missions. PRO-RR applies the PRO-X mission objectives to research reactor system design. This document serves as one of the foundational documents for the PRO-RR-Fuel System Design technical team by documenting current research reactor fuels usage. The PRO-RR-Fuel System Design technical team consists of subject matter experts from Argonne National Laboratory (Argonne) and Savannah River National Laboratory (SRNL). In order to determine the preferred fuel of use in upcoming research and test reactors to optimize proliferation resistance, performance, and safety, it is useful to assess the fuels that have been used in the past, or are currently in use. This report reviews the historical and current fuels used in research and test reactors to inform future fuel selection. Chapter 2 discusses the low-enriched uranium (LEU) fuels currently in use in terms of thermal power level and utilization of the reactor. Chapter 3 summarizes the fabrication processes for common fuel types. Chapter 4 discusses in detail the fuel types in use in research and test reactors. A review of the cladding types in use is presented in Chapter 5, and a historical review of research and test reactor fuel fabricators is presented in Chapter 6. The data collection strategy used the International Atomic Energy Agency (IAEA) research reactor database [1] as a starting point. Information on the fuel used was gathered on research reactors (other than critical assemblies) that were listed as operational, planned, or in temporary shutdown in the IAEA database. Data on the fuel type, geometry, enrichment, uranium loading, cladding type, and fabricator were collected for each of the reactors available in the public domain. Sources of data included conference papers, journal articles, and facility and fabricator websites. Data on research reactors operating on LEU fuels are presented in Appendix A, while Appendix B presents data collected on all reactors at the time of publication of this report. Appendix C presents data collected on reactors that were part of the M3 research and test reactor conversion program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Integrated Carbon Capture and Propane Oxidative Dehydrogenation for Olefins Production under Isothermal Conditions over Cr₂O₃-ZSM-5/CaO-CaZrO₃ Reactive Capture Materials

Converting captured CO₂ into fuels and marketable chemicals using bifunctional materials (BFMs) represents a promising sustainable solution. In this study, a BFM composed of a Cr₂O₃/ZSM-5 catalyst and a CaO-CaZrO₃ sorbent was developed and tested for the integrated CO₂-oxidative dehydrogenation of propane (CO₂-ODHP) in a fixed bed reactor. The CaO-CaZrO₃ sorbent, prepared using the citrate sol-gel method, demonstrated high CO₂ adsorption capacity and improved cyclic stability compared to pure CaO. The BFM achieved 23.5% propane conversion, 52.5% propylene selectivity, and 68.7% olefin selectivity at isothermal adsorption-reaction conditions (630-630 °C). Overall, this study provides insights into enhancing CO2 adsorption and developing materials for integrated CO2 capture and conversion.

Baamran, Khaled↗
DOE OSTI2024

3D high-fidelity automated neutronics guided optimization of fusion blanket designs

The compact Fusion Pilot Plant (FPP) is defined in the recent National Academies of Sciences, Engineering, and Medicine report as the next step of fusion energy demonstration with a $50$ MWe peak net electricity production, $Q_e$ greater than $1$, and at least $3$ hours of continuous operation. This fusion pilot plant will be a test bed enabling materials, designs, and fuel management assessment, and it will represent an engineering challenge because of its high-fusion power and compact design targets. Previous reactor data is limited to experiments operating in different design space ranges. Therefore, design iterations and assessments should rely on high-fidelity first-principle theoretical and computational models. The high-fidelity integrated modeling of the plasma is a fundamental part of fusion energy research. However, the whole device modeling is often neglected, utilizing low-fidelity, system-level analysis. Recently, the need for high-fidelity multi-physics modeling was recognized, resulting in a selection of integrated tools. Further, autonomous design optimization requires a streamlined framework that perturbs the design point, reruns the analysis, and examines the outputs. However, high-fidelity analysis requires complex geometry specification that is difficult to perturb. This work presents the parametric CAD generation tool TRACER and a new neutronic workflow. TRACER allows the perturbation of the geometry representation, creating geometry files ready for further analysis. The streamlined neutronic workflow allows efficient and accurate calculations. The two new tools coupled together were used to perform a 3D high-fidelity multi-objective, multi-input optimization of an "ARC Class" compact tokamak design. The workflow was driven by an optimization driver for full automation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Characterization of advanced nuclear materials under extreme environments

The performance of materials under extreme environments poses important fundamental questions about the behavior of condensed matter under far-from-equilibrium conditions. These conditions create challenges in materials design, testing and evaluation. One important aspect of advancing nuclear power is the discovery and investigation of new classes of materials able to withstand the harsh environments in nuclear reactors. However, investigating and understanding the performance of these materials requires sophisticated tools and characterization techniques and skills. Furthermore, the goal of this special issue is to present recent research on the characterization of the response of candidate advanced nuclear materials to extreme environments.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Comparative Critical Mass Calculations for NNL and ENDF/B-VIII.0 Zirconium Hydride Thermal Scattering Laws [Slides]

Zirconium Hydride (ZrH x ) is a moderator material for TRIGA reactors and SNAP-10A.The thermal scattering has the behavior of quantum oscillator favorable for moderator feedback. ZrH x occurs as a corrosion product in LWR. Previously new phase specific TSL evaluations were generated and submitted to ENDF/B- VIII.1. These include the effects of crystal structure on elastic scattering of Zr(ZrH x ), permit flexibility in stoichiometry for use in reactor physics calculations and fuel characterization, and expand the TSL material sub-library. ENDF/B-VIII.0 ZrH TSLs do not distinguish between material phase. Current work involves preliminary validation & verification testing on the effect of TSLs on critical mass and associated thermal flux.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Characterization of Aluminum Metal Powders for Plutonium-238 Program

As part of radioisotope thermoelectric generator (RTG) production for power sources in deep space exploration, Oak Ridge National Laboratory (ORNL) produces plutonium-238 ( 238 Pu). To produce 238 Pu, neptunium-237 ( 237 Np) targets are fabricated at ORNL and subsequently irradiated at the High Flux Isotope Reactor (HFIR) at ORNL and the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Specifically, current flowsheets utilize neptunium dioxide ( 237 NpO 2 ) targets. Aluminum (Al) powder is blended with 237 NpO 2 target material to improve thermal conductivity during irradiation. The current composition of 237 NpO 2 /Al pellets (i.e., cermets, or ceramic–metallics) is 20% 237 NpO 2 , 70% Al, and 10% void space. Al powders utilized by the 238 Pu program are high-fired under vacuum before blending into targets. The primary objective of this work is to measure the physical, chemical, and thermal properties of Al before and after the high-firing process. Properties of interest include the morphology, surface texture, particle size, crystal phase(s), and thermal conductivity of the material. These properties are measured with existing materials characterization equipment at ORNL, including powder x-ray diffraction (pXRD), scanning electron microscopy (SEM) with energy-dispersive spectroscopy (EDS), laser flash analysis (LFA), and thermogravimetric analysis (TGA). A secondary objective of this work is to compare the Al powders before and after the high-firing procedure to determine the effects of high-firing on chemical, physical, and thermal properties and to determine the efficacy of the high-firing process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Cooperative Research and Development Agreement (CRADA) between National Energy Technology Laboratory and Susteon, Inc. (Final Report)

NETL developed a nanostructured Fe catalyst capable of converting syngas to olefins. Initial laboratory testing demonstrated this technology possessed high reactivity (~70% conversion), good olefin selectivity (~40 weight %), and a long-performance lifetime (greater than 500 h on stream with no degradation). Under CRADA Agreement 0988, NETL and Susteon collaborated to evaluate the commercial potential of this catalyst technology by scaling up the catalyst synthesis and collecting reactivity data to characterize the mass and energy balance associated with utilizing this catalyst in a Fischer Tropsch to Olefins (FTO) reactor. Research under CRADA Agreement 0988 illustrated there are technical barriers associated with synthesizing the catalyst that the team was not able to overcome. Reactivity testing with this material illustrated that the catalyst deactivated in less than 50 hours of time on stream. Due to this technical barrier, the team did not meet its Go/No-Go decision target of creating a FTO process that produces at least a 25% reduction in olefin production cost compared to the current state-of the-art technologies. As a result, the team has stopped all work on CRADA Agreement 0988.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Mercury Modeling of TREAT Big-BUSTER Free Field Characterization Experiments

This report summarizes modeling of Big-BUSTER free field characterization (FFC) experiments at the Idaho National Laboratory (INL) Transient Reactor Test Facility (TREAT) reactor performed from December 20th, 2023, to February 22nd, 2024. TREAT is an air-cooled graphite moderated research reactor that can be used to study material response to neutron irradiation. The BUSTER (Broad Use Specimen Transient Experiment Rig) refers to a containment module for placing experiments in the reactor core. The Big-BUSTER is an analogous platform for a modified core configuration with a larger experimental cavity.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1973

Effect of nitrogen on tensile properties and structures of T-111 (tantalum, 8 percent tungsten, 2 percent hafnium) tubing

The effect of controlled nitrogen additions was evaluated on the mechanical properties of T-111 (Ta-8W-2Hf) fuel pin cladding material proposed for use in a lithium-cooled nuclear reactor concept. Additions of 80 to 1125 ppm nitrogen resulted in increased strengthening of T-111 tubular section test specimens at temperatures of 25 to 1200 C. Homogeneous distributions of up to 500 ppm nitrogen did not seriously decrease tensile ductility. Both single and two-phase microstructures, with hafnium nitride as the second phase, were evaluated in this study.

Buzzard, R. J.↗
NASA NTRS1992

Progress in SP-100 tribological coatings

The SP-100 Tribology Development program to investigate candidate coating materials for the SP-100 reactor is discussed. The SP-100 will operate at temperatures up to 1500 K in high vacuum. Background information, specific bearing requirements, the basis for coating material selection, testing methods, and initial results of the program are addressed.

Ring, Peter J.↗
DOE OSTI2023

The Innovation of I-Loop

The closing and decommissioning of Halden Boiling Water Reactor in Norway resulted in the loss of many nuclear irradiation testing facilities, notably those providing boiling water reactor (BWR) environments. To lighten the loss of such an important asset, the Advanced Test Reactor (ATR) is being called upon. The recent overhaul of the ATR’s core and the installation of the new Top Head Closure Plate now makes it possible to utilize the ATR’s I-loop locations for I-Loop Tube test facilities. Similar to the current in-pile tube facilities, these new I-Loop facilities will capitalize on the available space and neutron flux the ATR generates in this region of the core for fuels and materials testing. Current efforts are underway to add new PWR and BWR I-Loop Tube facilities to the I-Loop positions for these purposes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗