A study of thermal radiation absorption processes in gas core reactors
Analysis of thermal emission spectrum from uranium plasma in propellant region of gas core reactor
Celment, J. D.↗
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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
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Design of an external-fueled thermionic diode for in-pile testing.
Description of an external-fueled thermionic diode suitable for in-pile testing in a research reactor. The active electrode area is 94 sq cm. The 10-in. long, 1.5-in.-OD emitter body is tungsten 2% thoria. The fuel is contained in six 0.4-in.-diam holes equally spaced about the 0.5-in. central emitter hole. The collector is niobium-1% zirconium. The expected diode performance is 6 W/sq cm at 2000 K. In addition to following the constraints imposed by the in-pile testing and the electrically heated performance mapping prior to insertion in-pile, the diode will have end configurations prototypical of those anticipated for a flow-through, NaK-cooled, external-fuel thermionic reactor.
Ernst, D. M.↗
LVDT gage for fracture-toughness tests in liquid hydrogen
Linear-variable differential transformer replaces conventional resistance strain gages to measure crack-mouth-opening displacement. LVDT is superior in tests under liquid hydrogen, where boiling of hydrogen on resistive is suited to broad temperature range and hostile environments such as nuclear reactors.
Pierce, W. S.↗
Fluidized bed coal combustion reactor
A fluidized bed coal reactor includes a combination nozzle-injector ash-removal unit formed by a grid of closely spaced open channels, each containing a worm screw conveyor, which function as continuous ash removal troughs. A pressurized air-coal mixture is introduced below the unit and is injected through the elongated nozzles formed by the spaces between the channels. The ash build-up in the troughs protects the worm screw conveyors as does the cooling action of the injected mixture. The ash layer and the pressure from the injectors support a fluidized flame combustion zone above the grid which heats water in boiler tubes disposed within and/or above the combustion zone and/or within the walls of the reactor.
Moynihan, P. I.↗
Fabrication of light-weighted Si/SiC lidar mirrors
Small models of lightweight Si/SiC lidar mirrors have been fabricated via a scalable and cost-effective chemical vapor deposition (CVD) process. These mirrors are about 7.5 cm in diameter and consist of a Si-cladded SiC faceplate and lightweight structure made of SiC. The lightweight structure consists of an outer hexagonal cell with six triangular inner cells. The mirrors were polished to a figure better than 1/5 of a wave and a finish of better than 5A rms. The CVD route to fabrication of mirrors is fast and has the potential to yield several mirrors in a few weeks time from a signel reactor.
Goela, Jitendra S.↗
MHTGR Core Bypass Flow Patterns and Pressure Losses
The Modular High Temperature Gas-Cooled Reactor (M
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Modeling the Pore Formation Mechanism in UMo/AL Dispersion Fuel
In UMo/Al dispersion fuel meat, pores formed in the ILs or at IL-Al interfaces tend to increase in size with irradiation, potentially limiting performance of this fuel. There has been no universally accepted mechanism for the formation and growth of this type of pore. However, there is a consensus that the stress state determined by meat swelling and fission- induced creep is one of the determinants, and fission gas availability at the pore site is another. Five dispersion RERTR miniplates that have well defined irradiation conditions and PIE data were selected for examination. Meat swelling and pore volume were measured in each plate. ABAQUS finite element analysis (FEA) package was utilized to obtain the time-dependent evolution of mechanical states in the plates while matching the measured meat swelling and creep. Interpretation of these results give insights on how to model a failure function – a predictor for large pore formation – using variables such as meat swelling, interaction layer growth, stress, and creep. This model can be used for optimizing fuel design parameters to reach the desired goal: meeting high power and performance reactor demand.
Kim, Yeon Soo↗
Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario
The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that significantly exceeded QE limits for extended time periods for the low frequency STSBO sequence evaluated in this study. It is recognized that the core damage frequency (CDF) of the sequence evaluated in this scoping effort would be considerably lower if evaluations considered new FLEX equipment being installed by industry. Nevertheless, because of uncertainties in instrumentation response when exposed to conditions beyond QE limits and alternate challenges associated with different sequences that may impact sensor performance, it is recommended that additional evaluations of instrumentation performance be completed to provide confidence that operators have access to accurate, relevant, and timely information on the status of reactor systems for a broad range of challenges associated with risk important severe accident sequences.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
VTR Startup Fuel Paper for NFSM
A 2017 report issued by the Department of Energy Office of Nuclear Energy’s (DOE-NE) Nuclear Energy Advisory Committee (NEAC) discussed needs and requirements for a new test reactor in the U.S. The report included a recommendation to “proceed immediately with pre-conceptual design planning activities to support a new test reactor (including cost and schedule estimates).”
99 GENERAL AND MISCELLANEOUS↗
The chemistry of hydrogen in liquid-alkali- metal mixtures useful as nuclear reactor coolants. I.
Hydrogen solubility in eutectic sodium-potassium mixture, noting usefulness as nuclear reactor coolants, and dependence on pressure and temperature variations
EUTECTIC ALLOY↗
A REPORT: “HIGH-DENSITY U-MO/AL AND U3SI2/AL DISPERSION FUELS FOR HIGH-POWER RESEARCH REACTORS”
An introduction to a new report on high-density U-Mo/Al and U3Si2/Al dispersion fuels for high power research reactors that ANL is drafting is given in this paper. The main objective of this report is to produce a stand-alone data compilation that can be used by fuel designers, fuel performance analyzers, and fuel safety evaluators, as well as reactor operators. It is also intended to build a foundation for an eventual U-Mo/Al dispersion fuel qualification report that supports the licensing of this fuel for high-power research reactors. U-silicide dispersion fuel is also included in the report for potential applications beyond the ranges of the fuel licensed in NUREG-1313 (Safety Evaluation Report for U3Si2/Al). The contents and framework of the dispersion fuel report is described. A gap analysis provides a summary of the areas where sufficient data and information are not currently available.
Research reactor fuel, U-Si, U-Mo↗
The energy release and temperature field in the ultracold neutron source of the WWR-M reactor at the Petersburg Nuclear Physics Institute
Results of calculations of energy releases and temperature fields in the ultracold neutron source under design at the WWR-M reactor are presented. It is shown that, with the reactor power of 18 MW, the power of energy release in the 40-L volume of the source with superfluid helium will amount to 28.5 W, while 356 W will be released in a liquid-deuterium premoderator. The lead shield between the reactor core and the source reduces the radiative heat release by an order of magnitude. A thermal power of 22 kW is released in it, which is removed by passage of water. The distribution of temperatures in all components of the vacuum structure is presented, and the temperature does not exceed 100°C at full reactor power. The calculations performed make it possible to go to design of the source.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Investigation of Heterogeneous Reactions of C1ONO(sub 2), HC1, and HOC1 on Liquid Sulfuric Acid Surfaces
The heterogeneous reactions on liquid sulfuric acid surfaces have been studied using a fast flow-reactor coupled to a quadrupole mass spectrometer.
liquid sulfuric acid surfaces fast flow-reactor qu↗
Overview of the US Department of Energy Light Water Reactor Sustainability Program
The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Initial development of a method for correlating indentation test results to damage accumulation in high temperature structural materials
This report describes an experimental approach based on microindentation creep or relaxation tests for assessing the remaining creep life of a sample material. The objective for this experimental program is to complement the development of passively loaded in situ surveillance specimens. These surveillance specimens could be placed in future advanced reactors to assess the degradation of structural properties, primary creep and creep-fatigue life, when the material is exposed to actual reactor operating conditions. The specimens and the corresponding indentation test protocol could form the basis of a structural health monitoring program designed to ensure the safe continued operation of the component given the relatively unknown effect of the reactor coolant chemistry, radiation exposure, and other environmental effects on the material properties of the constituent structural materials.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Reactor production of Thorium-229
Limited availability of 229 Th for clinical applications of 213 Bi necessitates investigation of alternative production routes. In reactor production, 229 Th is produced from neutron transmutation of 226 Ra, 228 Ra, 227 Ac and 228 Th. Here, we evaluate irradiations of 226 Ra, 228 Ra, and 227 Ac targets at the ORNL High Flux Isotope Reactor.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
SAVANNAH RIVER SITE'S H-CANYON FACILITY: IMPACTS OF FOREIGN OBLIGATIONS ON SPECIAL NUCLEAR MATERIAL DISPOSITION
The US has a non-proliferation policy to receive foreign and domestic research reactor returns of spent fuel materials of US origin. These spent fuel materials are returned to the Department of Energy (DOE) and placed in storage in the L-area spent fuel basin at the Savannah River Site (SRS). The foreign research reactor returns fall subject to the 123 agreements for peaceful cooperation. These “123 agreements” are named after section 123 of the Atomic Energy Act of 1954 and govern the conditions of nuclear cooperation with foreign partners. The SRS management of these foreign obligations while planning material disposition paths can be a challenge.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Fail-safe reactivity compensation method for a nuclear reactor
The present invention relates generally to the field of compensation methods for nuclear reactors and, in particular to a method for fail-safe reactivity compensation in solution-type nuclear reactors. In one embodiment, the fail-safe reactivity compensation method of the present invention augments other control methods for a nuclear reactor. In still another embodiment, the fail-safe reactivity compensation method of the present invention permits one to control a nuclear reaction in a nuclear reactor through a method that does not rely on moving components into or out of a reactor core, nor does the method of the present invention rely on the constant repositioning of control rods within a nuclear reactor in order to maintain a critical state.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗