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DOE OSTI2012

The Development of A Human Systems Simulation Laboratory: Strategic Direction

The Human System Simulation Laboratory (HSSL) at the Idaho National Laboratory is one of few facilities of its kind that allows human factors researchers to evaluate various aspects of human performance and human system interaction for proposed reactor designs and upgrades. A basic system architecture, physical configuration and simulation capability were established to enable human factors researchers to support multiple, simultaneous simulations and also different power plant technologies. Although still evolving in terms of its technical and functional architecture, the HSSL is already proving its worth in supporting current and future nuclear industry needs for light water reactor sustainability and small modular reactors. The evolution of the HSSL is focused on continual physical and functional refinement to make it a fully equipped, reconfigurable facility where advanced research, testing and validation studies can be conducted on a wider range of reactor technologies. This requires the implementation of additional plant models to produce empirical research data on human performance with emerging human-system interaction technologies. Additional beneficiaries of this information include system designers and HRA practitioners. To ensure that results of control room crew studies will be generalizable to the existing and evolving fleet of US reactors, future expansion of the HSSL may also include other SMR plant models, plant-specific simulators and a generic plant model aligned to the current generation of pressurized water reactors (PWRs) and future advanced reactor designs. Collaboration with industry partners is also proving to be a vital component of the facility as this helps to establish a formal basis for current and future human performance experiments to support nuclear industry objectives. A long-range Program Plan has been developed for the HSSL to ensure that the facility will support not only the Department of Energy’s Light Water Reactor Sustainability Program, but also to provide human factors guidance for all future developments of the nuclear industry.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

Low Cost High-H 2 Syngas Production for Power and Liquid Fuels

This report summarizes the technical progress made of the research project entitled “Low Cost High-H2 Syngas Production for Power and Liquid Fuels,” under DOE Contract No. DE-FE-0011958. The period of performance was October 1, 2013 through July 30, 2015. The overall objectives of this project was to determine the technical and economic feasibility of a systems approach for producing high hydrogen syngas from coal with the potential to reduce significantly the cost of producing power, chemical-grade hydrogen or liquid fuels, with carbon capture to reduce the environmental impact of gasification. The project encompasses several areas of study and the results are summarized here. (1) Experimental work to determine the technical feasibility of a novel hybrid polymer/metal H2-membrane to recover pure H2 from a coal-derived syngas was done. This task was not successful. Membranes were synthesized and show impermeability of any gases at required conditions. The cause of this impermeability was most likely due to the densification of the porous polymer membrane support made from polybenzimidazole (PBI) at test temperatures above 250 °C. (2) Bench-scale experimental work was performed to extend GTI's current database on the University of California Sulfur Recovery Process-High Pressure (UCSRP-HP) and recently renamed Sulfur Removal and Recovery (SR2) process for syngas cleanup including removal of sulfur and other trace contaminants, such as, chlorides and ammonia. The SR2 process tests show >90% H2S conversion with outlet H2S concentrations less than 4 ppmv, and 80-90% ammonia and chloride removal with high mass transfer rates. (3) Techno-economic analyses (TEA) were done for the production of electric power, chemical-grade hydrogen and diesel fuels, from a mixture of coal- plus natural gas-derived syngas using the Aerojet Rocketdyne (AR) Advanced Compact coal gasifier and a natural gas partial oxidation reactor (POX) with SR2 technology. Due to the unsuccessful experimental results with the hybrid polymer/metal H2 membrane, a conventional CO2 capture (single-stage Selexol) and hydrogen purification (PSA) technologies were used in the appropriate cases. In all cases, the integrated system of Advanced Compact coal gasifier, non-catalytic natural gas partial oxidation, and SR2 multicontaminant removal with state-of-the-art auxiliary system provided a 5-25% cost advantage over the base line plants using GEE coal gasifier with conventional Selexol/Claus sulfur removal and recovery. These plants also produce 18-30% less CO2 than with the conventional coal gasification plants.

08 HYDROGEN↗

DOE OSTI2019

Development of Irradiation Test Devices for Transient Testing

The Transient Reactor Test facility (TREAT) resumed operations in 2017 in order to reclaim its crucial role in nuclear-heated fuel safety research. TREAT’s historic era of operation (1959-1994) was best known for integral-scale testing of large fuel specimens/bundles under postulated reactor plant accident conditions, but TREAT also supported smaller-scale phenomena identification tests that elucidated fundamental behaviors and paved the way for these integral-scale tests. Advances in modern computational capabilities and a resurgence of interest in novel reactor technologies have created an opportunity for emphasizing modernized science-based and separate effects tests once again at TREAT. An innovative approach to this type of testing has been developed to leverage minor radioactivity built-in during brief TREAT irradiations by arranging smaller fuel specimens in low activation hardware so that they can be easily extracted and shipped for post irradiation examination within weeks. This recently-established capability, termed the Minimal Activation Retrievable Capsule Holder (MARCH) irradiation vehicle system, includes capabilities for cost-effective simplified-environment testing of centimeter-scale fuel samples of various geometries, temperature-controlled irradiations of millimeter-size samples for lower-length-scale model development, liquid-metal bonded heat sink capsules for controlling transient temperature response in fuel rodlets, and an innovative approach to high throughput irradiation of transient sensors and instrumentation. The MARCH system’s capabilities will also set the foundation for fuel safety research performed in larger integral-scale test devices with coolant environments representing reactor plants. Based upon historic approaches, but modernized to meet current nuclear technology needs, these larger irradiation devices include flowing pressurized water (including the ability to depressurize to steam) as well liquid metal cooling loops for various fuel rod and small bundle specimens. Here, this paper describes the recently-established MARCH system and current trajectory to enabling advanced transient science with a suite of irradiation test devices.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2007

The ISS Water Processor Catalytic Reactor as a Post Processor for Advanced Water Reclamation Systems

Advanced water processors being developed for NASA s Exploration Initiative rely on phase change technologies and/or biological processes as the primary means of water reclamation. As a result of the phase change, volatile compounds will also be transported into the distillate product stream. The catalytic reactor assembly used in the International Space Station (ISS) water processor assembly, referred to as Volatile Removal Assembly (VRA), has demonstrated high efficiency oxidation of many of these volatile contaminants, such as low molecular weight alcohols and acetic acid, and is considered a viable post treatment system for all advanced water processors. To support this investigation, two ersatz solutions were defined to be used for further evaluation of the VRA. The first solution was developed as part of an internal research and development project at Hamilton Sundstrand (HS) and is based primarily on ISS experience related to the development of the VRA. The second ersatz solution was defined by NASA in support of a study contract to Hamilton Sundstrand to evaluate the VRA as a potential post processor for the Cascade Distillation system being developed by Honeywell. This second ersatz solution contains several low molecular weight alcohols, organic acids, and several inorganic species. A range of residence times, oxygen concentrations and operating temperatures have been studied with both ersatz solutions to provide addition performance capability of the VRA catalyst.

Nalette, Tim↗

DOE OSTI2024

Core Design Optimization of the Westinghouse Lead Fast Reactor

Westinghouse is pursuing an advanced Nuclear Power Plant design based on Lead Fast Reactor (LFR) technology for global commercialization. To achieve an optimal combination of key attributes, such as safety, sustainability, and economic competitiveness, Westinghouse and ANL partnered in developing and applying a formalized core design optimization strategy. An LFR analysis workflow was developed to automate a suite of reactor physics, fuels performance, safety, and economics simulations on a selected LFR concept. The workflow streamlines analysis of a wide range of LFR designs with different dimensions and fuel types to assess their viability and economic performance, significantly reducing human processing time and risks of processing errors. The LFR optimization exercise was defined, resulting in selection of the design constraints (geometric, neutronics, thermo-mechanical, safety, thermal-hydraulics, and economics) and performance metrics researched (minimization of both the fuels LCOE and the first core inventory cost). A total of 14 varied design parameters were considered, including assembly dimensions, coolant temperature, and enrichment distribution throughout the core. The LFR analysis workflow was connected to DAKOTA for sensitivity and optimization analyses. Due to the extremely large size of the potential LFR optimization solution space relative to the computing time required to characterize one LFR solution, a multi-stage optimization approach was proposed to breakdown the problem into several stages with more reasonable sizes. This optimization approach enabled finding various viable core solutions with different cost tradeoffs that were considered by Westinghouse and justify selection of a smaller core with multi-batch 2-year cycle length.

Stauff, Nicolas E.↗

DOE OSTI2012

Recommendations for NEAMS Engagement with the NRC

The vision of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program is to bring a new generation of analytic tools to the nuclear engineering community in order to facilitate students, faculty, industry and laboratory researchers in investigating advanced reactor and fuel cycle designs. Although primarily targeting at advanced nuclear technologies, it is anticipated that these new capabilities will also become interesting and useful to the nuclear regulator Consequently, the NEAMS program needs to engage with the Nuclear Regulatory Commission as the software is being developed to ensure that they are familiar with and ready to respond to this novel approach when the need arises. Through discussions between key NEAMS and NRC staff members, we recommend annual briefings to the Division of Systems Analysis in the NRC’s Office of Nuclear Regulatory Research.

Bernholdt, David E.↗

DOE OSTI2015

Carbide Coatings for Nickel Alloys, Graphite and Carbon/Carbon Composites to be used in Fluoride Salt Valves

The focus of this research was concerned with developing materials technology that supports the evolution of Generation IV Advanced High Temperature Reactor (AHTR) concepts. Specifically, we investigate refractory carbide coatings for 1) nickel alloys, and 2) commercial carbon-carbon composites (CCCs). Numerous compelling reasons have driven us to focus on carbon and carbide materials. First, unlike metals, the strength and modulus of CCCs increase with rising temperature. Secondly, graphite and carbon composites have been proven effective for resisting highly corrosive fluoride melts such as molten cryolite [Na₃AlF₆] at ~1000°C in aluminum reduction cells. Thirdly, graphite and carbide materials exhibit extraordinary radiation damage tolerance and stability up to 2000°C. Finally, carbides are thermodynamically more stable in liquid fluoride salt than the corresponding metals (i.e. Cr and Zr) found in nickel based alloys.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Online Monitoring of Induction Motors

The online monitoring of active components project, under the Advanced Instrumentation, Information, and Control Technologies Pathway of the Light Water Reactor Sustainability Program, researched diagnostic and prognostic models for alternating current induction motors (IM). Idaho National Laboratory (INL) worked with the Electric Power Research Institute (EPRI) to augment and revise the fault signatures previously implemented in the Asset Fault Signature Database of EPRI’s Fleet Wide Prognostic and Health Management (FW PHM) Suite software. Induction Motor diagnostic models were researched using the experimental data collected by Idaho State University. Prognostic models were explored in the set of literature and through a limited experiment with 40HP to seek the Remaining Useful Life Database of the FW PHM Suite.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Acceptance Test Data for BWXT Coated Particle Batch 93164A Defective IPyC Fraction and Pyrocarbon Anisotropy

Coated particle fuel batch J52O-16-93164 was produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used as demonstration production-scale coated particle fuel for other experiments. The tristructural-isotropic (TRISO) coatings were deposited in a 150-mm-diameter production-scale fluidizedbed chemical vapor deposition (CVD) furnace onto 425-μm-nominal-diameter spherical kernels from BWXT lot J52L-16-69316. Each kernel contained a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO) and was coated with four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (i.e., 93164A).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Acceptance Test Data for Candidate AGR-5/6/7 TRISO Particle Batches BWXT Coater Batches 93165 93172 Defective IPyC Fraction and Pyrocarbon Anisotropy

Coated particle fuel batches J52O-16-93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). Some of these batches may alternately be used as demonstration coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μmnominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

An Overview of Natural Gas Conversion Technologies for Co-Production of Hydrogen and Value-Added Solid Carbon Products

This report was prepared in response to the U.S. Department of Energy Fuel Cell Technologies Office Congressional Appropriation language to support research on carbon-free production of hydrogen using new chemical processes that utilize natural gas to produce solid carbon and hydrogen. The U.S. produces 9-10 million tons of hydrogen annually with more than 95% of the hydrogen produced by steam-methane reforming (SMR) of natural gas. SMR is attractive because of its high hydrogen yield; but it also converts the carbon to carbon dioxide. Non-oxidative thermal decomposition of methane to carbon and hydrogen is an alternative to SMR and produces CO 2 -free hydrogen. The produced carbon can be sold as a co-product, thus providing economic credit that reduces the delivered net cost of hydrogen. The combination of producing hydrogen with potentially valuable carbon byproducts has market value in that this allows greater flexibility to match the market prices of hydrogen and carbon. That is, the higher value product can subsidize the other in pricing decisions. In this report we highlight the relevant technologies reported in the literature—primarily thermochemical and plasma conversion processes—and recent research progress and commercial activities. Longstanding technical challenges include the high energetic requirements (e.g., high temperatures and/or electricity requirements) necessary for methane activation and, for some catalytic processes, the separation of solid carbon product from the spent catalyst. We assess current and new carbon product markets that could be served given technological advances, and we discuss technical barriers and potential areas of research to address these needs. We provide preliminary economic analysis for these processes and compare to other emerging (e.g., electrolysis) and conventional (e.g., SMR) processes for hydrogen production. The overarching conclusion of this study is that the cost of hydrogen can be potentially reduced to target levels of $2/kg with the co-production and sale of a sufficiently high-value carbon product. Technological advances are required to understand the reaction conditions and design reactor systems that can achieve high yields of the select carbon products and segregate or separate the high-value carbon products, and optimize the production process for both hydrogen and carbon.

30 DIRECT ENERGY CONVERSION↗

DOE OSTI2021

Fabrication and Characterization of Single-Particle Compacts

The programmatic objective of the advanced low enriched uranium (aLEU) fuel Project within the Office of Nonproliferation Research and Development (NA-22) is to advance the development of nuclear reactor designs, fuel materials, and fabrication technologies capable of meeting a number of challenging customer requirements, including reactor stability, extended lifetime, and power density without refueling, while using only low-enriched uranium (i.e., less than 20% enrichment of 235 U). The tandem requirements of delivering sufficient power density and maintaining fuel performance over the course of a multidecade fuel lifetime tightly constrain the field of potential fuel options. Monolithic fuel forms (e.g., fuel pellets) provide maximized density of fissile material, but they may face challenges in maintaining their structure and performance over the long reactor lifetimes envisioned in this program. Dispersion fuel forms with an inert matrix provide enhanced long-term stability but sacrifice fissile material density, and they may not provide sufficient power density over the reactor lifetime with the limit of low enrichment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Multiscale and Machine Learning Modeling for Additive Manufacturing

Additive manufacturing (AM) techniques provide the opportunity to simultaneously design new materials and components with complex structures in less time, enabling faster material developments. Even though compositionally similar, the texture of the materials produced by such techniques is significantly different from conventionally manufactured materials. Additively manufactured materials produces highly heterogeneous microstructure within a single build. Such variations in the microstructure make qualifying AM products challenging for extreme environment applications. Understanding the AM process and its influence on the materials’ microstructures/properties is paramount for evaluating the workability and performance of the manufactured materials. The performance of AM materials for advanced nuclear reactor applications is of interest to the Advanced Materials and Manufacturing Technologies (AMMT) program under the Department of Energy Office of Nuclear Energy. Hence, considering the microstructural variabilities in the AM products and their impact on the performance of the material, it is important to correlate the process conditions to the final product and establish a process-structure-property- performance (PSPP) correlation for AM materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1993

Nuclear propulsion technology advanced fuels technology

Viewgraphs on advanced fuels technology are presented. Topics covered include: nuclear thermal propulsion reactor and fuel requirements; propulsion efficiency and temperature; uranium fuel compounds; melting point experiments; fabrication techniques; and sintered microspheres.

Stark, Walter A., Jr.↗

NASA NTRS1993

Space Nuclear Thermal Propulsion (SNTP) tests

Viewgraphs on the space nuclear thermal propulsion (SNTP) program are presented. The objective of the research is to develop advanced nuclear thermal propulsion (NTP) technology based on the particle bed reactor concept. A strong philosophical commitment exists in the industry/national laboratory team to emphasize testing in development activities. Nuclear testing currently underway to support development of SNTP technology is addressed.

Allen, George C.↗

NASA NTRSDate not supplied

Lunar Helium-3: Mining Concepts, Extraction Research, and Potential ISRU Synergies

This paper describes various methods and designs specifically proposed for mining helium-3, and summarizes the methods, results and potential implications of recent research on helium extraction from lunar regolith simulant. This paper also touches on the potential synergy of lunar propellant production with helium-3 mining and some of the recent advancements in fusion technology related to future helium-3 fueled fusion reactors.

Lunar↗

NASA NTRSDate not supplied

Lunar Helium-3: Mining Concepts, Extraction Research, and Potential ISRU Synergies

Lunar↗

DOE OSTI2015

Prognostic and health management of active assets in nuclear power plants

This study presents the development of diagnostic and prognostic capabilities for active assets in nuclear power plants (NPPs). The research was performed under the Advanced Instrumentation, Information, and Control Technologies Pathway of the Light Water Reactor Sustainability Program. Idaho National Laboratory researched, developed, implemented, and demonstrated diagnostic and prognostic models for generator step-up transformers (GSUs). The Fleet-Wide Prognostic and Health Management (FW-PHM) Suite software developed by the Electric Power Research Institute was used to perform diagnosis and prognosis. As part of the research activity, Idaho National Laboratory implemented 22 GSU diagnostic models in the Asset Fault Signature Database and two wellestablished GSU prognostic models for the paper winding insulation in the Remaining Useful Life Database of the FW-PHM Suite. The implemented models along with a simulated fault data stream were used to evaluate the diagnostic and prognostic capabilities of the FW-PHM Suite. Knowledge of the operating condition of plant asset gained from diagnosis and prognosis is critical for the safe, productive, and economical long-term operation of the current fleet of NPPs. This research addresses some of the gaps in the current state of technology development and enables effective application of diagnostics and prognostics to nuclear plant assets.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗