Use of tungsten-184 as a thermal flux monitor at high temperatures
Tungsten-184 used as thermal flux monitor at high temperatures in Plum Brook reactor
Ball, L. L.↗
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International Symposium on Combustion, 14th, Pennsylvania State University, University Park, Pa., August 20-25, 1972, Proceedings.
Some elementary steps for the formation and destruction of nitric oxide in the H-N-O system, high-temperature oxidation of hydrogen by nitrous oxide in shock waves, and high-temperature fast-flow reactor studies of metal-atom oxidation kinetics are among the topics covered in papers concerned with elementary reactions in combustion. Other areas covered include oscillatory, turbulent, and supersonic combustion; flames in furnaces and combustors; and pollutant formation and destruction in flames. Individual items are announced in this issue.
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Applications of plasma core reactors to terrestrial energy systems
Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-
Latham, T. S.↗
Simplified Heat-Source/Thermionic Converter
Radiation coupling of heat from heat-source cylinder to converter cylinder through vacuum gap eliminates need for high-temperature electrical insulators between reactor heat pipes and thermionic converters. In addition no radiatior heat pipe is necessary because collectors of thermionic converters from which excess heat must be removed radiate directly to space. New design concept is also applicable to terrestrial and non-nuclear thermionic power supplies.
Shimada, K.↗
Modeling and Testing of Non-Nuclear, Highpower Simulated Nuclear Thermal Rocket Reactor Elements
When the President offered his new vision for space exploration in January of 2004, he said, "Our third goal is to return to the moon by 2020, as the launching point for missions beyond," and, "With the experience and knowledge gained on the moon, we will then be ready to take the next steps of space exploration: human missions to Mars and to worlds beyond." A human mission to Mars implies the need to move large payloads as rapidly as possible, in an efficient and cost-effective manner. Furthermore, with the scientific advancements possible with Project Prometheus and its Jupiter Icy Moons Orbiter (JIMO), (these use electric propulsion), there is a renewed interest in deep space exploration propulsion systems. According to many mission analyses, nuclear thermal propulsion (NTP), with its relatively high thrust and high specific impulse, is a serious candidate for such missions. Nuclear rockets utilize fission energy to heat a reactor core to very high temperatures. Hydrogen gas flowing through the core then becomes superheated and exits the engine at very high exhaust velocities. The combination of temperature and low molecular weight results in an engine with specific impulses above 900 seconds. This is almost twice the performance of the LOX/LH2 space shuttle engines, and the impact of this performance would be to reduce the trip time of a manned Mars mission from the 2.5 years, possible with chemical engines, to about 12-14 months.
Kirk, Daniel R.↗
Experimental Studies of NaK in a Simulated Space Environment
Space fission power systems are being developed at the National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) with a short term goal of building a full scale, non-nuclear, Technology Demonstration Unit (TDU) test at NASA's Glenn Research Center. Due to the geometric constraints, mass restrictions, and fairly high temperatures associated with space reactors, liquid metals are typically used as the primary coolant. A eutectic mixture of sodium (22 percent) and potassium (78 percent), or NaK, has been chosen as the coolant for the TDU with a total system capacity of approximately 55 L. NaK, like all alkali metals, is very reactive, and warrants certain safety considerations. To adequately examine the risk associated with the personnel, facility, and test hardware during a potential NaK leak in the large scale TDU test, a small scale experiment was performed in which NaK was released in a thermal vacuum chamber under controlled conditions. The study focused on detecting NaK leaks in the vacuum environment as well as the molecular flow of the NaK vapor. This paper reflects the work completed during the NaK experiment and provides results and discussion relative to the findings.
Gibons, Marc↗
Silicon Carbide Microsystems Technology for Extreme Environments: Opportunities for Nuclear Applications
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silicon carbide↗
Final Technical Report
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22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Analysis of Flow Area in the Axial Cross-Vessel Annular Duct Following a Breach of the Primary Helium Boundary of an HTGR
The lock-exchange flow phenomenon has been heavily studied for decades, including the lock-exchange phenomenon that takes place during an air ingress accident in a High-Temperature Gas-cooled Reactor (HTGR) during a break on the axial cross-vessel duct. The present study aims to characterize the gas dynamics in the cavity region of an HTGR, particularly the axial cross-vessel outer duct region. This study computes the velocity and flow area of the heavy and light currents when a break on the primary helium boundary occurs. This analysis is supported using experimental data from experiments conducted in a 1/20th length-scale HTGR. The following study summarizes the conditions of the experiment, the approach adopted to predict the flow area, and future work.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Best-estimate Modeling of the High Temperature Test Facility with RELAP5-3D
Prismatic HTGRs are a concept of interest for near-term deployment. While plenty of validation data exist for standalone neutronics or multiphysics modeling, the availability of integral effects thermal hydraulics validation data is more limited. The High Temperature Test Facility provides such data and is used as the basis for the High-Temperature Gas-Cooled Reactor Thermal Hydraulics Benchmark. This presentation presents results of best-estimate modeling of HTTF experiments PG-27 and PG-29 using the RELAP5-3D ring model developed at Idaho National Laboratory.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Buffer-IPyC separation process in TRISO fuel particles simulated with Bison code
During High Temperature Gas-cooled Reactor (HTGR) operation, tristructural isotropic (TRISO) coated-particle fuel undergoes irradiation-induced changes in morphology and thermomechanical properties. Experimental results from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program show, among other things, the mechanism of gap formation between the buffer and inner pyrolytic carbon (IPyC) layers, which could be explored further via computational simulations using the Bison code. Two simulation models were developed, the debonding restricted model, where no gap formation between buffer and IPyC layers is permitted, and the debonding enabled model, where the gap between those layers is created. The inputs of the simulated models are based on the irradiation conditions from the AGR-1 experiment. The research included simulations on spherical and aspherical fuel types. Under the specific temperatures and fluences of the AGR-1 irradiation experiment, and the Bison simulations, it was concluded that the most common scenario is a gap formation along the buffer-IPyC interface, while the least possible scenario is the situation where there is no gap formation at the buffer-IPyC junction. The computational results confirmed that the sphericity of the fuel influences the thickness of the gap that occurs at the buffer-IPyC junction, in a way that with increasing aspect ratio the gap thickness increases. The results obtained for spherical and aspherical fuel are nearly identical. Finally, performed simulations match conclusions observed from the AGR-1 experiment, which as such shows that the Bison code is a good computational method for simulating the TRISO fuel. Future simulations will include the validation of performed research and comparison of the results between Bison and PARFUME codes.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
RAVEN Template for Dynamic Representativity Analysis of the High Temperature Test Facility
These slides present a walkthrough of the template that has been developed for using RAVEN to perform representativity analysis using models of the High Temperature Test Facility and the General Atomics Modular High Temperature Gas-cooled Reactor. The presentation provides participants in the HTTF benchmark with a walkthrough on how to use RAVEN for their sensitivity analysis and how to read results from the MHTGR-350 to perform representativity
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Assessment of buffer-IPyC thermomechanical debonding behavior using new experimental strength data in BISON
TRIstructural ISOtropic (TRISO) fuel is a nuclear fuel commonly used in High Temperature Gas-cooled Reactors (HTGRs). A single sub-millimeter-diameter TRISO fuel particle consists of a spherical fuel kernel surrounded by four coating layers: a low-density pyrocarbon buffer layer, an inner pyrolytic carbon (IPyC) layer, a silicon carbide (SiC) layer, and an outer pyrolytic carbon (OPyC) layer. The kernel is commonly made of UO2 or a mixture of uranium carbide and uranium oxide (UCO). During reactor operation, the TRISO coating layers are subjected to irradiation-induced dimensional changes and the associated thermomechanical behavior of each layer. One of the observed behaviors is gap formation between the buffer and IPyC layer due to the porous buffer’s irradiation-induced shrinkage exceeding that of the IPyC layer. Not all irradiated particles will experience buffer-IPyC gap formation. The debonding may be partial, or it may be nearly total. However, from post-irradiation examination of UCO TRISO fuels irradiated as part of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program, it was concluded that partial buffer-IPyC debonding was the most common type of buffer-IPyC interaction. To predict TRISO thermomechanical performance, multi-physics models have been built that are being continually updated and refined. The BISON code is a finite element-based nuclear fuel performance code that may be used for 1D, 2D, and 3D TRISO particle simulations. This code is used to calculate fuel temperature, kernel swelling, buffer densification, thermal and irradiation creep, fracture, and fission gas production and release. One of the recent additions to the BISON code is the ability to model the process of layer debonding. This paper will focus on the simulation results of the improved BISON debonding model that will utilize updated strengths measured from irradiated AGR TRISO fuel particles. The new experimental strength data from micromechanical tests of irradiated TRISO fuel samples were exercised in the BISON simulations and compared to baseline strength data to assess their applicability in the models. This also includes updated buffer-IPyC bond strengths to simulate layer delamination. Based on current experimental observations it is noted that the buffer-IPyC separation occurs not exactly at the junction of these two layers, but more on the side of the buffer layer. That observation is also implemented in the TRISO interface debonding model. This improved modeling approach using experimental strength data to characterize buffer-IPyC debonding and its potential subsequent cracking will be presented in the paper along with comparisons to available experimental observations.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Advanced Coolant Enclosure Solutions for Micro Gas Cooled Reactors with Enhanced Efficiency and Safety
Argonne’s advanced high-pressure, high-temperature gas enclosures solutions were developed and tested for Light Water Reactor (LWR) fuel applications wherein pressure, temperature, fuel and working fluid are substantially different when compared to the operating conditions typical of gas cooled reactors (GCR). For LWR applications the coolant sleeve pressure exerted on the inner and outer walls of the sleeve may be equalized by pressurizing the fuel rod (e.g., with helium gas) to compensate for the coolant high pressure exerted on the outer walls of the sleeve. This results in a lower differential pressure. Micro-GCRs are reactor systems operated at elevated temperatures and pressure with gas as a working fluid. HolosGen’s Holos-Quad represent a type of Micro-GCR wherein the core is sleeved to segregate the fuel and other components from the working fluid.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Hydrogen Recovery by Methane Pyrolysis to Elemental Carbon
Use of a Sabatier reactor to recover the oxygen from the carbon dioxide exhaled by the crew on the International Space Station has been limited by the loss of the hydrogen contained in the methane it generates. Maximizing the oxygen recovered requires the hydrogen to be recovered from the methane product and recycled back to the Sabatier reactor. We describe the use of a tailored methane pyrolysis reactor to completely recover this hydrogen. The carbon-containing byproduct is elemental carbon, which is generated in the form of easily handled, non-sooty material that may have various uses. The process of creating this tailored carbon vapor deposition process involved exploration of the effects of temperature, pressure, substrate design and other variables to develop a high yield process that cleanly generates the desired products. Reaction kinetics and kinetics modelling were used to specify the temperature, pressure and reactor volume required to achieve the target conversion and to assure that the final average density was as high as possible. Reactor design included the selection of materials that will survive the high temperatures and environment in the pyrolysis reactor, and thermal modeling to achieve the required temperatures with minimum power consumption. The successful construction and demonstration of a brassboard prototype will allow the results of the chemical, thermal and mechanical models to be validated and should provide a useful alternative for a completely closed loop ECLS system. Integration of this technology with state-of-the-art (SOA) Sabatier hardware on ISS requires a complete understanding of the effects of impurities in the product hydrogen on the Sabatier catalyst. SOA Sabatier catalyst was evaluated over short and long-term exposure to anticipated contaminants to identify effects.
Yates, Stephen F.↗
Hydrogen Recovery by Methane Pyrolysis to Elemental Carbon
Use of a Sabatier reactor to recover the oxygen from the carbon dioxide exhaled by the crew on the International Space Station has been limited by the loss of the hydrogen contained in the methane it generates. Maximizing the oxygen recovered requires the hydrogen to be recovered from the methane product and recycled back to the Sabatier reactor. We describe the use of a tailored methane pyrolysis reactor to completely recover this hydrogen. The carbon-containing byproduct is elemental carbon, which is generated in the form of easily handled, non-sooty material that may have various uses. The process of creating this tailored carbon vapor deposition process involved exploration of the effects of temperature, pressure, substrate design and other variables to develop a high yield process that cleanly generates the desired products. Reaction kinetics and kinetics modelling were used to specify the temperature, pressure and reactor volume required to achieve the target conversion and to assure that the final average density was as high as possible. Reactor design included the selection of materials that will survive the high temperatures and environment in the pyrolysis reactor, and thermal modeling to achieve the required temperatures with minimum power consumption. The successful construction and demonstration of a brassboard prototype will allow the results of the chemical, thermal and mechanical models to be validated and should provide a useful alternative for a completely closed loop ECLS system. Integration of this technology with state-of-the-art (SOA) Sabatier hardware on ISS requires a complete understanding of the effects of impurities in the product hydrogen on the Sabatier catalyst. SOA Sabatier catalyst was evaluated over short and long-term exposure to anticipated contaminants to identify effects.
Yates, Stephen F.↗
Yttrium Hydride Post Irradiation Examination Plan
The overarching goal of the Department of Energy Office of Nuclear Energy (DOE-NE) microreactor program is to develop technologies for the deployment of civilian microreactors by stakeholders1. Microreactors are expressed as advanced transportable nuclear reactors operating at low power (<20MWth) but high temperatures (>600°C), as well as plug-and-play and inherently safe designs. One prerequisite of a microreactor is the compactness, so that a truck can transport the reactor under safe conditions with the current road infrastructure1,2. The compactness of these reactors likely can be attainable by use of solid components for the essentials of the nuclear core, such as fuel, core heat removal components, reflectors, and moderators. Among these essentials, where fuel enrichment must remain < 20% to meet High Assay Low Enrichment Uranium criteria (HALEU), the largest contribution to the compactness is offered by use of solid moderators which benefit from light atomic weight elements, such as hydrogen, carbon, and beryllium2. Among these, hydrogen-bearing materials, such as metal hydrides, are superior to other options from the lowest critical mass standpoint. Noting that, factors other than critical mass should be considered for a specific reactor design. Yttrium- or zirconium-based metal hydrides have been down-selected due to their neutronic performance. In addition to the neutronic perspective, maintaining hydrogen within the metal hydride is important at the high operating temperatures proposed by advanced reactors. Yttrium hydride (YHx) is, therefore, a proposed moderator material that offers better hydrogen retention at higher operating temperatures than zirconium hydrides due to higher retention and thermal stability of hydrogen in the metal3. The irradiated materials properties of metal hydrides, in this case YHx, must be assessed for the qualification of these moderators. Material testing and inspection processes must illustrate that the effects of dimensional and property changes on thermophysical and mechanical properties do not cause any significant changes on the microreactor safety, and the moderating power is maintained within design limits. Thus, the effect of irradiation on the thermophysical and mechanical properties must be determined. This post-irradiation examination (PIE) plan specifically aims to determine these properties for YHx following Advanced Test Reactor (ATR) irradiation.
99 GENERAL AND MISCELLANEOUS↗
PLN-6268 Yttrium Hydride Post Irradiation Examination Plan
The overarching goal of the Department of Energy Office of Nuclear Energy (DOE-NE) microreactor program is to develop technologies for the deployment of civilian microreactors by stakeholders1. Microreactors are expressed as advanced transportable nuclear reactors operating at low power (<20MWth) but high temperatures (>600°C), as well as plug-and-play and inherently safe designs. One prerequisite of a microreactor is the compactness, so that a truck can transport the reactor under safe conditions with the current road infrastructure1,2. The compactness of these reactors likely can be attainable by use of solid components for the essentials of the nuclear core, such as fuel, core heat removal components, reflectors, and moderators. Among these essentials, where fuel enrichment must remain < 20% to meet High Assay Low Enrichment Uranium criteria (HALEU), the largest contribution to the compactness is offered by use of solid moderators which benefit from light atomic weight elements, such as hydrogen, carbon, and beryllium2. Among these, hydrogen-bearing materials, such as metal hydrides, are superior to other options from the lowest critical mass standpoint. Noting that, factors other than critical mass should be considered for a specific reactor design. Yttrium- or zirconium-based metal hydrides have been down-selected due to their neutronic performance. In addition to the neutronic perspective, maintaining hydrogen within the metal hydride is important at the high operating temperatures proposed by advanced reactors. Yttrium hydride (YHx) is, therefore, a proposed moderator material that offers better hydrogen retention at higher operating temperatures than zirconium hydrides due to higher retention and thermal stability of hydrogen in the metal3. The irradiated materials properties of metal hydrides, in this case YHx, must be assessed for the qualification of these moderators. Material testing and inspection processes must illustrate that the effects of dimensional and property changes on thermophysical and mechanical properties do not cause any significant changes on the microreactor safety, and the moderating power is maintained within design limits. Thus, the effect of irradiation on the thermophysical and mechanical properties must be determined. This post-irradiation examination (PIE) plan specifically aims to determine these properties for YHx following Advanced Test Reactor (ATR) irradiation.
99 GENERAL AND MISCELLANEOUS↗