Cavity reactor gas-core critical experiment
Critical experiment for cavity reactor using gaseous uranium hexafluoride core configuration
Hyland, R. E.↗
SEARCH · Search NASA
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.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Nuclear thermal design of high temperature gas-cooled reactor with SiC/C mixed matrix fuel compacts - 18410
Japan Atomic Energy Agency (JAEA) has started R and D on SiC/C mixed matrix fuel element of high temperature gas-cooled reactors (HTGRs) to improve oxidation resistance of fuel. Nuclear thermal design of HTGR with SiC/C mixed matrix fuel compacts was carried out as a part of above R and Ds. Nuclear thermal design was carried out based on a small sized HTGR for developing countries, HTR50S, which thermal power is 50 MW and burn-up period is 2 years. Maximum enrichment of uranium is set to be 10 wt%, because coated fuel particles with 10 wt% uranium have been fabricated in Japan. Numbers of kinds of enrichment and burnable poisons (BPs) were set to be same as those of original HTR50S (3 and 2, respectively). We succeeded in nuclear thermal design of a small sized HTGR which performance was equivalent to original HTR50S, with SiC/C mixed matrix fuel compacts. Based on nuclear thermal design, intactness of coated fuel particles was evaluated to against internal pressure increase by burnup. It was shown that intactness was kept for burnup condition of HTR50S. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
PDC Modifications for Analysis of Gas-Cooled Reactors with Direct Helium Brayton Cycle
Capabilities of the Plant Dynamics Code (PDC) have been extended to allow steady-state and transient simulation of graphite-cooled reactors with direct helium Brayton cycle. On the cycle side, the most significant code modification is the addition of helium properties, in the format required by the code’s equations. Since the code was already formulated to handle more complex real gas properties, adding helium as working fluid that behaves like ideal gas was fairly straightforward. A reactor module was added to PDC to simulate a reactor cooled by the working fluid of the Brayton cycle. Two options are supported: channel type, typical for graphite gas-cooled reactors, and pin type, typical for light-water and liquid metal-cooled reactors. The reactor module is an extension of the electrical heater model and simulates heat deposition in the fuel and transfer of this heat from the fuel to the coolant through the matrix and tube materials. The new reactor module becomes the third option in PDC for modeling heat addition to the cycle, besides previously modeled heat addition heat exchanger and electrical heater. In addition to those changes, other minor code modifications and improvements were introduced during the work of expanding PDC to modeling of gas-cooled reactors. These modifications are summarized in the last chapter of this report.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Preliminary results of AGC-4 irradiation in the advanced test reactor and design of AGC-5 - 18469
The U.S. Department of Energy Advanced Reactor Technologies Program will irradiate up to six nuclear graphite creep experiments in Idaho National Laboratory Advanced Test Reactor (ATR). The graphite experiments are being irradiated over an approximate 8-year period to support development of a graphite irradiation performance database on new nuclear grade graphites now available for use in high-temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the very high-temperature gas reactor and other future gas reactors. Each of the experiments consist of a single capsule that contains six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load and the other half of specimens not subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of the diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with online temperature and compressive load monitoring and control. Also, samples are taken of sweep gas effluent to measure any oxidation or off-gassing that may occur during initial startup of the experiment. The first experiment Advanced Graphite Capsule (AGCJ-I started irradiation in September 2009 and completed in January 2011; the second experiment AGC-2 started irradiation in April 2011 and completed in May 2012; the third experiment AGC-3 started irradiation in late November 2012 and completed in April 2014; and AGC-4 is currently being irradiated in ATR. This paper will briefly discuss preliminary irradiation results from the AGC-4 experiment and design of AGC-3. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Thermal Fluid Modeling Approaches in SAM for High Temperature Gas-Cooled Reactor Applications
This paper provides an overview of the unique modeling and simulation challenges and needs in the system and safety analysis of two common types of High Temperature Gas-cooled Reactor (HTGR) designs. The challenges are associated with the complex core geometry configurations and the change of the dominant heat removal mechanisms between normal operating conditions and decay heat removal transients. This requires that simulation tools utilize models with reasonable length scales and capture both the localized heat transfer and the core-scale heat transfer at the same time. To meet such analysis needs, different methodologies were developed. In conclusion, this includes adopting existing methods and further developing/improving them, as well as proposing innovative methods that leverage advanced computational frameworks such as the Multiphysics Object Oriented Simulation Environment (MOOSE).
HTGR↗
Coupled SAM/Griffin Model of a Reference Pebble Bed High-Temperature Gas Cooled Reactor for Multi-Physics Simulations
An effort has been dedicated to developing a reference model for multi-physics coupled simulation of the pebble bed high-temperature gas-cooled reactor (PB-HTGR) with SAM and Griffin computer codes for safety analysis and licensing purpose. The reference problem of a prototypical reactor design serves as the foundation for the U.S. NRC (Nuclear Regulatory Commission) to verify the adequacy of computer codes and evaluation models for specific reactor types. In this work, a SAM model of the HTR-PM reactor has been developed based on publicly available design information and the multi-dimensional Pronghorn model developed by Idaho National Laboratory. The SAM HTR-PM model includes a multi-dimensional core region and 0- D/1-D fluid components. Additionally, a simplified air RCCS loop is modeled for decay heat removal. The Griffin model is based on a recent work by Idaho National Laboratory. The coupling between the models is achieved through the MOOSE MultiApp system. Both steady-state and transient scenarios were simulated to demonstrate the coupled model’s capability for multi-physics simulations.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
An analysis of the flow field near the fuel injection location in a gas core reactor.
An analytical study is presented which shows the effects of large energy release and the concurrent high acceleration of inner stream fluid on the coaxial flow field in a gas core reactor. The governing equations include the assumptions of only radial radiative transport of energy represented as an energy diffusion term in the Euler equations. The method of integral relations is used to obtain the numerical solution. Results show that the rapidly accelerating, heat generating inner stream actually shrinks in radius as it expands axially.
Weinstein, H.↗
First of a Kind (FOAK) Nuclear Demonstration Readiness Projects: Advancement of High Temperature Gas Cooled Reactors for Grid-Scale Applications: Xe-100 Conceptual Design and Risk-Informed, Performance Based Licensing
Abstract for First of a Kind (FOAK) Nuclear Demonstration Readiness Projects: Advancement of High Temperature Gas Cooled Reactors.
Mulder, Eben↗
Gas core reactors for actinide transmutation
The preliminary design of a uranium hexafluoride actinide transmutation reactor to convert long-lived actinide wastes to shorter-lived fission product wastes was analyzed. It is shown that externally moderated gas core reactors are ideal radiators. They provide an abundant supply of thermal neutrons and are insensitive to composition changes in the blanket. For the present reactor, an initial load of 6 metric tons of actinides is loaded. This is equivalent to the quantity produced by 300 LWR-years of operation. At the beginning, the core produces 2000 MWt while the blanket generates only 239 MWt. After four years of irradiation, the actinide mass is reduced to 3.9 metric tonnes. During this time, the blanket is becoming more fissile and its power rapidly approaches 1600 MWt. At the end of four years, continuous refueling of actinides is carried out and the actinide mass is held constant. Equilibrium is essentially achieved at the end of eight years. At equilibrium, the core is producing 1400 MWt and the blanket 1600 MWt. At this power level, the actinide destruction rate is equal to the production rate from 32 LWRs.
Clement, J. D.↗
Reduced-Order Modeling of Multigroup Neutron Cross Sections for High-Temperature Gas-cooled Reactors
Abstract – Deterministic neutronics calculations rely on multigroup neutron cross section libraries, which consist of databases of tabulated values, used to calculate the neutron cross sections through multivariate linear interpolation. However, interpolation of the multidimensional cross section data becomes memory inefficient and time consuming as the number of tabulations increases, significantly slowing down the neutronics calculation, especially in the case of microscopic cross section libraries where every isotope (on the order of hundreds) has its own set of specific reactions and cross sections. In order to address this challenge, this work constructs efficient and robust reduced-order models (ROMs) of the multi-group cross sections to support the Griffin simulation of high-temperature gas-cooled reactors (HTGRs). The first part of the study investigates the linearity of the multigroup cross section data across isotopes, reaction types, and energy groups on pre-generated datasets for the purpose of dimensionality reduction. Secondly, a down-selection of ROM techniques is presented on representative classical machine learning (ML) techniques, including variants of linear regression, kernel-based methods, tree-based algorithms, and artificial neural networks. The selection criteria jointly consider the memory efficiency, predictive accuracy, prediction speed, and scalability in comparison to the multidimensional interpolation. Among all the ML techniques, deep neural networks (DNNs) have proven to be the best selection with sufficient accuracy, high robustness, good memory efficiency, great scalability, and superior flexibility. DNNs have been trained for all isotopes in this work and systematic Griffin testing is ongoing to ensure the feasibility of this ROM technique for predicting cross section and reducing memory requirements without a significant sacrifice in computational performance.
42 - ENGINEERING↗
Identification and Characterization of Thermal Fluid Phenomena Associated with Selected Operating/Accident Scenarios in Modular High Temperature Gas-cooled Reactors
Dominant phenomena are identified and characterized for the loss-of-forced-convection (LOFC) scenario, the depressurized loss-of-forced-convection (DLOFC) scenario, and the moisture ingress scenario with relief valves available scenario. These scenarios are considered for the modular high temperature gas-cooled reactor (MHTGR) baseline design—with both a prismatic core and a pebble-bed core. The objectives of this effort, described in the above paragraph, were achieved by following a methodology identified by the U.S. Nuclear Regulatory Commission (NRC) and recommended in their Regulatory Guide 1.203. The methodology employed is described. For the above scenarios, baseline phenomena identification and ranking tables (PIRTs) generated by a NRC-sponsored effort in 2008 for the LOFC and DLOFC scenarios and a baseline PIRT generated by a Department of Energy-sponsored effort in 2011 by their Next Generation Nuclear Plant (NGNP) Moisture Ingress Assessment Committee were used to identify the dominant phenomena and existing levels-of-knowledge. Ten dominant phenomena (with only medium or low levels-of-knowledge) were identified for the LOFC and DLOFC scenarios including: (1) Inlet plenum stratification and plumes, (2) core coolant flow distribution, (3) radiant heat transfer from the top of the core to the upper vessel head, (4) reactor cavity air circulation and heat transfer, (5) reactor cavity cooling system behavior, (6) decay heat distributions, (7) core effective thermal conductivity, (8) molecular diffusion in air ingress DLOFC, (9) duct exchange flow, (10) phenomena that affect cavity gas composition for air-ingress. Five dominant phenomena, with high levels of knowledge, were identified for the moisture-ingress scenario with relief valves available including: (1) flow through break, (2) moisture level in the primary, (3) transport to moisture monitor, (4) moisture monitor instrument response, and (5) heat removal by shutdown cooling system. These phenomena are discussed and characterized. A listing of experiments—both completed and ongoing—is given with a view toward including data useful for further more detailed examination of the dominant phenomena listed in the previous paragraph. Note that only data for the LOFC and DLOFC scenarios are listed—since the moisture ingress scenario with relief valves available –have been identified as having adequate levels-of-knowledge by the PIRT committee. Finally, a preliminary computational fluid dynamics (CFD) calculation of inlet plenum stratification and plumes is described. Techniques employed for this calculation are typical for preliminary scoping calculations performed to validate numeric tools.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor
Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.
Bragg-Sitton, S. M.↗
Beyond the classical kinetic model for chronic graphite oxidation by moisture in high temperature gas-cooled reactors
Four grades of nuclear graphite with various microstructures were subjected to accelerated oxidation tests in helium with traces of moisture and hydrogen in order to evaluate the effects of chronic oxidation on graphite components in high temperature gas cooled reactors. Kinetic analysis showed that the Langmuir-Hinshelwood (LH) model cannot consistently reproduce all results. In particular, at high temperatures and water partial pressures oxidation was always faster than the LH model predicts, with stronger deviations for superfine grain graphite than for medium grain grades. It was also found empirically that the apparent reaction order for water has a sigmoid-type variation with temperature which follows the integral Boltzmann distribution function. This suggests that the apparent activation with temperature of graphite reactive sites that causes deviations from the LH model is rooted in specific structural and electronic properties of surface sites on graphite. A semi-global kinetic model was proposed, whereby the classical LH model was modified with a temperature-dependent reaction order for water. The new Boltzmann-enhanced model (BLH) was shown to consistently predict experimental oxidation rates over large ranges of temperature (800-1100 oC) and partial pressures of water (3-1200 Pa) and hydrogen (0-300 Pa), not only for the four grades of graphite but also for the historic grade H-451. The BLH model offers as more reliable input for modeling the chemical environment effects during the life-time operation of new grades of graphite in advanced nuclear reactors operating at high and very high temperatures.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model
Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.
Kazeminezhad, F.↗
Modeling of a Generic Pebble Bed High-temperature Gas-cooled Reactor (PB-HTGR) with SAM
This report presents the modeling of the core of a generic pebble-bed reactor (PBR) at the system level using the System Analysis Module (SAM) code. This work is an extension of a previous work by the authors (Ooi et al. (2021)) that used the so-called 2-D ring model approach to model the PBMR-400. With the new approach, the pebble bed of the reactor is modeled with multiple PBCoreChannel components with spherical heat structures which allows the code to calculate thermal fluid parameters with built-in closure relations. The new core-channel approach is an improvement to the 2-D ring model approach as it does not introduce geometric distortions to the model and thus reduces the uncertainties of the predictions. In addition to thermal fluid simulations, point kinetics (PKE) are included to the model. Simulations are performed under a steady-state normal operation condition and a load-following transient scenario. This particular transient scenario is chosen as it tests both the thermal fluid and neutronics aspects of the model. The predicted results from both the steady-state and transient scenarios are compared with the results by Stew- art et al. (2021) who performed similar simulations with a Griffin-Pronghorn coupled tool. Despite the differences between the codes, with SAM being a system-analysis code and Pronghorn being a porous-medium code, both sets of results compare favorably. The over- all profiles and trends of the predicted temperatures and reactivities from the SAM and Griffin/Pronghorn simulations are similar, with some differences in their predicted values. The first part of the report covers the significance of a relatively fast-running approach that is capable of modeling the pebble bed reactor at the system-level while simultaneously capturing the radial thermal behavior of the core. Then, the modeling approach used in this work is discussed in details. Lastly, the results and comparisons with the Griffin/Pronghorn simulation are presented.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Design and optimization of flexible decoupled high-temperature gas-cooled reactor plants with thermal energy storage
Advanced nuclear power plants are well-positioned for future zero-carbon grids, however, the need for flexible power generation will be required over the traditional emphasis on baseload generation for meeting historical demands. To achieve such flexibility, this work examines viable configurations for coupling nuclear energy production with thermal energy storage. Previous designs on nuclear-thermal energy storage configurations for advanced reactor designs, which utilized reactor steam as the heat source for charging the thermal energy storage, are restricted by the heat diversion ratio and efficiency losses, thus their impacts can be limited. In this context, this study proposes configurations for fully decoupling the nuclear reactor from the power cycle and positioning the storage as an intermediate loop, thereby achieving an unconstrained heat diversion ratio and improved efficiency. Compared with a standard high-temperature gas-cooled reactor’s power cycle, steady-state thermodynamic modeling and dispatch optimizations quantify the benefits of a steam reheat cycle within the fully decoupled thermal energy system to separate the plant cycle from the high-pressure primary side. These benefits are further detailed, compatible with required high-temperature and high-pressure conditions, through (1) open-source dynamic transient models that examine the impact of off-design operation on the systems, (2) the investigation of components design and costing and finally (3) sizing and dispatch optimization. The fully-decoupled design achieves a cycle efficiency of 43.1%, an enhancement over the vendor’s standard efficiency of 42.2% (Xe-100 design). Here, the proposed design offers strengthened physical barriers from the nuclear island as well as superior operational flexibility and power boosting. Dispatch optimization and market analysis reveal that thermal energy storage size is highly dependent on the peak patterns of electricity prices and the minimum generation level constraint imposed on the balance of plant. Evaluation of off-design operation demonstrates that the full decoupling design with the suggested fail-safe control mechanisms ensures a minimal impact on reactor parameters, even during rapid power ramping.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Gas core reactor concepts and technology - Issues and baseline strategy
Results of a research program including phenomenological studies, conceptual design, and systems analysis of a series of gaseous/vapor fissile fuel driven engines for space power platforms and for thermal and electric propulsion are reviewed. It is noted that gas and vapor phase reactors provide the path for minimum mass in orbit and trip times, with a specific impulse from 1020 sec at the lowest technololgical risk to 5200 sec at the highest technological risk. The discussion covers various configurations of gas core reactors and critical technologies and the nuclear vapor thermal rocket engine.
Diaz, Nils J.↗