Regulatory Development Support- Holbrook SOW-18109
This SOW will support the review of the NGNP PIRTs for the TRISO and High-Temperature Gas Reactor programs.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
SEARCH · Engineering Papers
Results for “GAS REACTOR”
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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Performance of a catalytic reactor at simulated gas turbine combustor operating conditions
The performance of a catalytic reactor 12 cm in diameter and 17 cm long was evaluated at simulated gas turbine combustor operating conditions using premixed propane and air. Inlet temperatures of 600 and 800 K, pressures of 3 and 6 atm, and reference velocities of 9 to 30 m/s were tested. Data were taken for equivalence ratios as high as 0.43. The operating range was limited on the low-temperature side by very poor efficiency; the minimum exit temperature for good performance ranged from 1400 to 1600 K depending on inlet conditions. As exit temperatures were raised above this minimum, emissions of unburned hydrocarbons decreased, carbon monoxide emissions became generally less than 1 g CO/kg fuel, and nitrogen oxides were less than about 0.1 g NO2/kg fuel.
Anderson, D. N.↗
Coupling of Pronghorn and RELAP-7 for a Pebble Bed Reactor
High temperature gas cooled reactors (HTGR) are a candidate for timely Gen-IV reactor technology deployment because of high technology readiness and walk-away safety. Among HTGRs, pebble bed reactors (PBRs) have attractive features such as low excess reactivity and online refueling. Pebble bed reactors pose unique challenges to analysts and reactor designers such as continuous burnup distribution depending on pebble motion and recirculation, radiative heat transfer across a variety of gas-filled gaps, and long design basis transients such as pressurized and depressurized loss of forced circulation. Modeling and simulation is essential for both the PBR’s safety case and design process. In order to verify and validate the new generation codes the Nuclear Energy Agency (NEA) Data bank provide a set of benchmarks data together with solutions calculated by the participants using the state of the art codes of that time. An important milestone to test the new PBR simulation codes is the OECD NEA PBMR-400 benchmark which includes thermal hydraulic and neutron kinetic standalone exercises as well as coupled exercises and transients scenarios. In this work, the reactor multiphysics code MAMMOTH and the thermal hydraulics code Pronghorn, both developed by the Idaho National Laboratory (INL) within the multiphysics object-oriented simulation environment (MOOSE), have been used to solve Phase 1 exercises 1 and 2 of the PBMR-400 benchmark. The steady state results are in agreement with the other participants’ solutions demonstrating the adequacy of MAMMOTH and Pronghorn for simulating PBRs.
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Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input
A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Exploratory study of several advanced nuclear-MHD power plant systems.
In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.
Williams, J. R.↗
[Research Seminar] HTGR R&D and Deployment in the US: Status and Challenges
Slides prepared for the research seminar at universities of the research progress funded by the Advanced Reactor Technologies (ART) - Gas Cooled Reactors (GCR) program.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Pressure Separation and Gas Flows in a Prototype Vacuum Pumped Solar Thermochemical Reactor.
Abstract not provided.
Ermanoski, Ivan↗
Integrated Water Gas Shift Membrane Reactors Utilizing Novel, Non Precious Metal Mixed Matrix Membrane
Nanoparticles of zeolitic imidazolate frameworks and other related hybrid materials were prepared by modifying published synthesis procedures by introducing bases, changing stoichiometric ratios, or adjusting reaction conditions. These materials were stable at temperatures >300 °C and were compatible with the polymer matrices used to prepare mixed- matrix membranes (MMMs). MMMs tested at 300 °C exhibited a >30 fold increase in permeability, compared to those measured at 35 °C, while maintaining H 2 /CO 2 selectivity. Measurements at high pressure (up to 30 atm) and high temperature (up to 300 °C) resulted in an increase in gas flux across the membrane with retention of selectivity. No variations in permeability were observed at high pressures at either 35 or 300 °C. CO 2 -induced plasticization was not observed for Matrimid®, VTEC, and PBI polymers or their MMMs at 30 atm and 300 °C. Membrane surface modification by cross-linking with ethanol diamine resulted in an increase in H 2 /CO 2 selectivity at 35 °C. Spectrometric analysis showed that the cross-linking was effective to temperatures <150 °C. At higher temperatures, the cross-linked membranes exhibit a H 2 /CO 2 selectivity similar to the uncross-linked polymer. Performance of the polybenzimidazole (PBI) hollow fibers prepared at Santa Fe Science and Technology (SFST, Inc.) showed increased flux and selectivity at 300 °C, which is comparable to a flat PBI membrane. A water-gas shift reactor has been built and currently being optimized for testing under DOE conditions.
36 MATERIALS SCIENCE↗
GIF Gas Cooled Fast Reactor: Proliferation Resistance and Physical Protection White Paper
This white paper represents the status of Proliferation Resistance and Physical Protection (PR&PP) characteristics for the Gas-cooled Fast reactor (GFR) reference designs selected by the Generation IV International Forum (GIF) GFR System Steering Committee (SSC). The intent is to generate preliminary information about the PR&PP features of the GFR reactor technology and to provide insights for optimizing their PR&PP performance for the benefit of GFR system designers. It updates the GFR analysis published in the 2011 report “Proliferation Resistance and Physical Protection of the Six Generation IV Nuclear Energy Systems”, prepared Jointly by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) and the System Steering Committees and provisional System Steering Committees of the Generation IV International Forum, taking into account the evolution of both the systems, the GIF R&D activities, and an increased understanding of the PR&PP features. The white paper, prepared jointly by the GIF PRPPWG and the GIF GFR SSC, follows the high-level paradigm of the GIF PR&PP Evaluation Methodology to investigate the PR&PP features of the GIF GFR 2400 MWth reference design. The ALLEGRO reactor is also described. The EM2 and HEN MHR reactor are mentioned. An overview of fuel cycle for the GFR reference design and for the ALLEGRO reactor are provided. For PR, the document analyses and discusses the proliferation resistance aspects in terms of robustness against State-based threats associated with diversion of materials, misuse of facilities, breakout scenarios, and production in clandestine facilities. Similarly, for PP, the document discusses the robustness against theft of material and sabotage by non-State actors. The document follows a common template adopted by all the white papers in the updated series.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
AGR-5/6/7 Fuel Fabrication, Process Documentation, and Disposition
The need for U. S. energy self-sufficiency dictates the development of new low-cost, environmentally-safe forms of energy generation, hydrogen production and industrial heat systems using high temperature gas-cooled reactor (HTGR) technology. Efficient fabrication of ultra-high quality, coated particle fuel is a central requirement for implementing gas reactor technology. Babcock & Wilcox Nuclear Operations Group-Lynchburg (B&W NOG-L) is uniquely qualified in the U.S. with the institutional knowledge, access to equipment, and a Nuclear Regulatory Commission (NRC) license to produce tristructural isotropic (TRISO) coated particle nuclear fuel above 5 wt% 235U. This Statement of Work (SOW) defines the Subcontractor, B&W NOG-L and identifies the work scope to perform Advanced Gas Reactor (AGR)-5/6/7 fuel fabrication.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Initial Assessment of Environmental Effects on SiC/SiC Composites in Helium-Cooled Nuclear Systems
Ceramic matrix composites, in particular SiC/SiC composites, are currently being evaluated for use at high temperatures and high radiation fields in advanced concept reactors, including high-temperature gas cooled reactors (HTGR), very high temperature reactors (VHTR), gas fast reactors (GFR), sodium fast reactors (SFR) and fluoride-cooled high temperature reactors (FHR). SiC/SiC composites have many desirable properties for high temperature nuclear applications, including excellent thermal and mechanical properties and radiation resistance. However, their behavior in contact with the chemical environment in normal operation conditions and in off-normal accident conditions scenarios still needs to be evaluated. This report summarizes the information available in the literature on the chemical reactivity of SiC/SiC composites and of their components in contact with the helium coolant used in HTGR, VHTR and GFR designs. Under normal operating conditions, ultra-high purity helium will have chemically controlled impurities (water, oxygen, carbon dioxide, carbon monoxide, methane, hydrogen) that will create a slightly oxidizing gas environment. Little is known from direct experiments on the reactivity of third generation (nuclear grade) SiC/SiC composites in contact with low concentrations of water or oxygen in inert gas, at high temperature. However, there is ample information regarding the oxidation behavior of SiC/SiC composites in dry and moist air at high temperatures. This information is reviewed first in the next chapters. The emphasis is placed on the improvement in material oxidation, thermal, and mechanical properties during three stages of development of SiC fibers and at least two stages of development of the fiber/matrix interphase. The chemical stability of SiC/SiC composites in contact with oxygen or steam at temperatures that may develop in off-normal reactor conditions supports the conclusion that most advanced composites (also known as nuclear grade SiC/SiC composites) have the chemical resistance that would allow them maintain mechanical properties at temperatures up to 1200 – 1300 °C in the extreme conditions of an air or water ingress accident scenario. Further research is needed to assess the long-term stability of advanced SiC/SiC composites in inert gas (helium) in presence of very low concentrations (traces) of water and oxygen at the temperatures of normal operation of helium-cooled reactors.
ceramic matrix composites↗
AGR-5/6/7 Fuel Fabrication, Process Documentation, and Disposition Planning
The need for U. S. energy self-sufficiency dictates the development of new low-cost, environmentally-safe forms of energy generation, hydrogen production and industrial heat systems using high temperature gas-cooled reactor (HTGR) technology. Efficient fabrication of ultra-high quality, coated particle fuel is a central requirement for implementing gas reactor technology. Babcock & Wilcox Nuclear Operations Group-Lynchburg (B&W NOG-L) is uniquely qualified in the U.S. with the institutional knowledge, access to equipment, and a Nuclear Regulatory Commission (NRC) license to produce tristructural isotropic (TRISO) coated particle nuclear fuel above 5 wt% 235U. This Statement of Work (SOW) defines the Subcontractor, B&W NOG-L and identifies the work scope to perform Advanced Gas Reactor (AGR)-5/6/7 fuel fabrication, process documentation, and material disposition planning.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
AGR-5/6/7 Fuel Fabrication, Process Documentation, and Disposition
The need for U. S. energy self-sufficiency dictates the development of new low-cost, environmentally-safe forms of energy generation, hydrogen production and industrial heat systems using high temperature gas-cooled reactor (HTGR) technology. Efficient fabrication of ultra-high quality, coated particle fuel is a central requirement for implementing gas reactor technology. Babcock & Wilcox Nuclear Operations Group-Lynchburg (B&W NOG-L) is uniquely qualified in the U.S. with the institutional knowledge, access to equipment, and a Nuclear Regulatory Commission (NRC) license to produce tristructural isotropic (TRISO) coated particle nuclear fuel above 5 wt% 235U. This Statement of Work (SOW) defines the Subcontractor, B&W NOG-L and identifies the work scope to perform Advanced Gas Reactor (AGR)-5/6/7 fuel fabrication, process documentation, and material disposition planning.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Gas core rocket reactors - A new look
Feasibility analysis of open-cycle gas-core nuclear rocket engine
Ragsdale, R. G.↗
Simulations of the High Temperature Test Facility using SAM
Under the support from the U.S. Department of Energy Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, an effort is being pursued to support the modeling and simulation needs of high-temperature gas-cooled reactor (HTGR) technology development. There is a particular need for advanced modeling and simulation methods and tools to predict thermal-fluid behavior in the nuclear reactor primary system during safety-related transients. This report focuses on one such activity related to HTGR: developing a model of the High Temperature Test Facility (HTTF) at Oregon State University using the system-level code SAM, and using the model to understand thermal response behavior in the facility. Note this activity is coordinated with the DOE-NE’s Advanced Reactor Technology Gas-Cooled Reactor Program
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
RELAP5-3D Modeling in the OECD-NEA HTTF Benchmark
Prismatic gas-cooled reactors are a technologically mature reactor concept of interest for near- to mid-term deployment. To accelerate the deployment of these reactors, the DOE's Advanced Reactor Technologies Gas-Cooled Reactor campaign is spearheading a thermal hydraulics code validation benchmark based on the High Temperature Test Facility at Oregon State University. This presentation provides an introduction to that benchmark and an overview of some of the RELAP5-3D validation activities stemming from that benchmark.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
HTGR Validation: NEUP Survey and Database Database Development for HTGR Thermal-Fluid Experiments
This set of slides is aimed to improve access to the High-Temperature Gas-cooled Reactor (HTGR) validation data and optimize the return on the significant investment made by DOE. Supported by the Advanced Reactor Technologies (ART) Gas-Cooled Reactor (GCR) program. Slides include information from FY2009 to FY2023, there are in total 35 DOE NEUP projects focusing on the thermal-fluid experiments related with High-Temperature Gas-cooled Reactor (HTGR), producing a large amount of high-quality validation data. NEUP Survey and Database Development for HTGR Thermal-Fluid Experiments is distributed at universities and has not been disseminated to the HTGR community. More collaborating with university PIs and refining the HTGR phenomena summary chart continuously is expected in the future.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
HTGR Validation: NEUP Survey and Database - Data Reporting Standard for HTGR Thermal-Fluid Experiments
Since 2009, the U.S. Department of Energy (DOE) Office of Nuclear Energy's Nuclear Energy University Program (NEUP) has been at the forefront of nuclear research, specifically concentrating on advancing high-temperature gas-cooled reactor (HTGR) technologies. By Fiscal Year 2024, NEUP has authorized 36 projects dedicated to HTGR research, each contributing significantly to the enhancement of our understanding of this technology. The outcomes of these diverse projects have been disseminated through final NEUP reports, peer-reviewed journal articles, and presentations at academic conferences, forming a comprehensive tapestry of knowledge. Despite the substantial value of these findings, their dissemination has been fragmented, posing challenges for accessibility to researchers and policymakers and leading to underutilization of DOE investments. Recognizing this critical gap and its potential consequences for the future of nuclear research, the Advanced Reactor Technologies (ART) Gas-Cooled Reactor (GCR) program conducted an extensive survey of completed and ongoing HTGR NEUP projects. This survey enabled the compilation of crucial data, resulting in the development of a specialized public-access database tailored for computational fluid dynamics and system code validation, specifically designed for HTGR applications. However, the data collection process revealed a significant challenge in central data organization due to individual researchers from different institutes employing varying logics and preferences for recording and documenting experimental data. Consequently, an urgent need has been identified to establish a standardized reporting format for HTGR experimental projects. Addressing this issue is essential for enhancing collaboration, maximizing the impact of DOE investments, and ensuring the seamless advancement of HTGR technologies in nuclear research.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗