Engineering Papers⌕ Search

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Thermal Hydraulics Validation Activities in the ART-GCR Campaign

The Advanced Reactor Technologies - Gas Cooled Reactor Campaign (ART-GCR) is part of the United States Department of Energy - Office of Nuclear Energy's Advanced Reactor Technologies program aimed at developing a High Temperature Gas-cooled Reactor (HTGR) which will offer enhancements in safety and efficiency. The Design, Methods, and Validation arm of the ART-GCR program provides experience and advanced tools for HTGR design and analysis. The Methods and Validation arm is developing data to validate predictions of decay heat removal through an extensive experimental campaign the Natural Convection Shutdown Heat Removal Test Facility at Argonne National Laboratory. In-core thermal hydraulics validation is being facilitated through the OECD/NEA's Thermal Hydraulic Code Validation Benchmark for High Temperature Gas-Cooled Reactors using HTTF Data. That benchmark contains code-to-code and code-to-data comparisons for 3 sets of HTGR thermal hydraulics phenomena. This ongoing benchmark activity has 16 participants from 9 countries developing leading insights into HTGR thermal hydraulics code validation. The ART-GCR campaign is also working on the development of state-of-the-art MOOSE-based models of the High Temperature Engineering Test Reactor (HTTR). Through the OECD/NEA's HTTR Loss of Forced Cooling (LOFC) program, the ART-GCR campaign is developing a multiphysics transient code validation benchmark based on 3 HTTR experiments that will provide an opportunity for international collaboration on multiphysics code validation.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

CFD Simulations of Lower Plenum Mixing

Review of model development and validation performed in the Advanced Reactor Technologies (ART) program for thermal mixing at the outlet of High Temperature Gas Reactors (HTGRs). Understanding the mixing that occurs in the lower plenum in an HTGR is necessary to facilitate design improvements and to perform reactor safety analysis. Numerical models are one possible approach to gain a better understanding of mixing in the lower plenum. Given the complexity of the geometry and the intense mixing present, it is important to perform validation of numerical models. Three models have been developed during FY2025: a porous media with Pronghorn, a Reynolds Averaged Navier Stokes (RANS) with STAR-CCM+, and a Large Eddy Simulation (LES) with NekRS. The reference facility is a scaled-down version of the lower plenum of the High Temperature Gas-Cooled Reactor - Pebble-bed Module (HTR-PM) demonstration reactor. Preliminary results of the porous media and the RANS shows general good agreement against experimental benchmark data. Future work will leverage high-fidelity results obtained through LES to guide model selection and improvements to the lower-fidelity models, with particular attention to the Pronghorn porous media.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Quantity of 135 I released from the AGR-1, AGR-2, and AGR-3/4 experiments and discovery of 131 I at the FPMS traps during the AGR-3/4 experiment

A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tristructural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The HPGe detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135m Xe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135 I. About 2.5 hours after irradiation (ten 15.3 minute 135m Xe half lives) the directly produced 135m Xe has decayed and only the longer lived 135 I remains as a source. Decay systematics dictate that 135m Xe will be in secular equilibrium with its 135 I parent, such that its production rate very nearly equals the decay rate of the parent, and its concentration in the flowing gas stream will appear to decay with the parent half life. This equilibrium condition enables the determination of the amount of 135 I released from the fuel particles by measurement of the 135m Xe at the FPM following reactor shutdown. In this paper, the 135 I released will be reported and compared to similar releases for noble gases as well as the unexpected finding of 131 I deposition from intentional impure gas injection into capsule 11 of experiment AGR 3/4.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Neutronic Design and Analysis of the Holos-Quad Concept

HolosGen, LLC is proposing a highly innovative micro-reactor concept targeting both civilian and military applications. It consists of an advanced helium-cooled gas reactor using a turbojet-type turbine and compressor to achieve a highly condensed reactor that fits into a commercial ISO container. The Holos-Quad concept being considered for civilian applications will generate 22MWt, using four Subcritical Power Modules (SPMs) that fit into one 40-foot ISO container. It is a high-temperature gas-cooled reactor concept using TRISO fuel distributed in graphite hexagonal blocks, cooled with helium gas at 7MPa and a high outlet temperature. In FY2019, HolosGen, LLC received an ARPA-E MEITNER grant to team up with the Argonne National Laboratory (ANL) design team to demonstrate the feasibility of the Holos-Quad concept in terms of neutronics. In particular, the evaluation study of the Holos design features and alternative options for reactivity control was considered. Through these tasks, the ANL design team has been working on demonstrating the feasibility of the Holos-Quad concept in terms of neutronics to help advance the core design of the Holos-Quad. A rigorous design approach was employed for the Holos-Quad in order to ensure that the input space is fully investigated, and that the best solutions are considered. The first step of this approach consisted of properly defining the Holos-Quad design problem by identifying the competitive design objectives: its economic performance is maximized with longer lifetime and its transportability to remote locations is maximized with reduced weight. The design problem also required to identify the operational constraints (in terms of thermal hydraulics, reactivity feedback and shutdown margins) and the input parameters that could be varied. In the second step, a sensitivity analysis was performed to enable preliminary investigation of the input space to identify the correlations among input and output variables. It allowed, for instance, to find input parameters that have relatively poor impact on the core performance and that could be removed from the optimization problem to accelerate its convergence. This sensitivity analysis showed that the design problem defined is highly constrained since only 117 solutions out of 1,000 simulated cores met all the design constraints. The design optimization was performed in the third step employing a genetic algorithm to effectively explore this highly constrained input space and find global optimal solutions. More than 2,500 cores were fully evaluated to converge to a Pareto Frontier of optimum solutions providing best compromise between the life-time target and the total core weight, as shown in Figure 1. The main benefit of this approach was to reduce the human effort and to be able to automatically re-run the optimization problem when changing the design problem. The multi-objective optimization identified various core solutions that would be optimum solutions for different types of applications.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Seismic isolation: A pathway to standardized advanced nuclear reactors

Standardizing advanced nuclear reactors is a pathway to substantially reducing their overnight capital cost and achieving parity with other power sources, including renewables and fossil fuels. The seismic load case has thwarted standardization of nuclear power plants because site-specific seismic hazard and local near-surface geology has triggered soil-structure-interaction analysis, design, equipment qualification, regulatory review, and licensing, ensuring that each build is different. To achieve standardized or site-independent certified advanced reactor designs, the impact of the seismic load case on the engineering and construction cost and time must be substantially mitigated. Seismic isolation is a mature technology that has been used for more than 30 years in non-nuclear sectors to substantially reduce earthquake demands in buildings and other infrastructure. In this paper, seismic isolation is used to enable standardization of advanced reactor designs, aimed at the complete re-use of a site-independent, certified design and repeated procurement of safety-class equipment. A pathway to standardized designs using seismic isolation is demonstrated for two fundamentally different advanced reactors: a molten salt reactor and a high temperature gas reactor. Each reactor building is equipped with three specialized pieces of safety-class equipment, namely, a reactor vessel, a steam generator, and a control rod drive mechanism housing that is attached to the reactor head. Analysis is performed per ASCE and ASME standards to design the buildings and the equipment for two base conditions: conventional (fixed base) and base isolated. The impact of the seismic load case is characterized for the reinforced concrete walls in the buildings and for the equipment, measured using vessel wall thickness and horizontal accelerations. Here the analysis results show that the fixed-base buildings, designed for a site of low seismic hazard (peak ground acceleration, PGA = 0.15 g) could be constructed at a site of much greater seismic hazard (PGA = 0.7 g) if seismic base isolation is employed. Importantly, the scope of the site-specific analysis, design, and qualification would be limited to the seismic isolators and the isolated substructure, drastically reducing plant-specific engineering, review, and licensing, and time to construction start. Regulatory challenges and opportunities with standardized reactor designs are identified.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

STEM/EELS Study of Fission Product Transport in Neutron Irradiated TRISO Fuel Particles (NSUF report)

The objectives of this project are to study the distribution, composition and structure of fission products in neutron irradiated SiC layer of the TRISO particles, using electron microscopy including scanning transmission electron microscopy (STEM), energy dispersive x-ray spectroscopy (EDS), electron energy loss spectroscopy (EELS), high-resolution transmission electron microscopy (HRTEM) and precession electron diffraction (PED), in an effort to enhance the understanding of distribution, composition, structure and transport mechanisms of fission products. In the past 40 years, numerous studies, including reactor experiments, out-of-pile experiments, and simulations, have been performed to investigate the fission product transport behavior and mechanisms. Grain boundary (GB) diffusion, neutron-induced diffusion, catalytic Pd-assisted transport, and vapor-phase migration all have been suggested as possible governing transport mechanisms. The previous studies, however, have not been able to reproduce the transport behavior of fission products and the mechanisms remain poorly understood. This project is anticipated to contribute to enhancing the understanding of the mechanisms of fission product transport in irradiated SiC layer of the TRISO particles. Thorough understanding of fission product transport mechanisms in the SiC layer will improve fuel modeling and design and better SiC layers and TRISO coated particles can be designed to enhance retention of fission products and reduce their release. Eventually, TRISO coated particles with improved performance can be fabricated, and such advancement will expedite the qualifying and licensing processes for TRISO coated particles. Hence, this research is contributing to achieving the objectives of DOE-NE to develop the next generation nuclear reactors, particularly high-temperature gas reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Micro Reactor Instrumentation and Control FY2019 Report

Micro-reactors (i.e., very small transportable or mobile nuclear reactors with a capacity less than 20 MWt) are being developed to supply heat and power for various applications in remote areas, military installations, emergency operations, humanitarian missions, and disaster relief zones. A wide variety of reactor types are under consideration, including sodium-cooled fast reactors, molten salt reactors, light water reactors, very-high-temperature gas reactors, and heat pipe reactors. These miniaturized transportable reactor designs remain largely untested and unproven. System and component testing are needed to demonstrate design safety and system robustness, reliability and efficiency.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Mechanistic Source Term Considerations for Advanced Non-LWRs

This report is a functional review of the radionuclide containment strategies of fluoride-salt-cooled high temperature reactor (FHR), molten salt reactor (IVISR) and high temperature gas reactor (HTGR) systems. This analysis serves as a starting point for further, more in-depth analyses geared towards identifying phenomenological gaps that still exist, preventing the creation of a mechanistic source term for these reactor types. As background information to this review, an overview of how a mechanistic source term is created and used for consequence assessment necessary for licensing is provided. How mechanistic source term is used within the LMP is also provided. Third, the characteristics of non-LWR mechanistic source terms are examined This report does not assess the viability of any software system for use with advanced reactor designs, but instead covers system function requirements. Future work within the Nuclear Energy Advanced Modeling and Simulations (NEAMS) program will address such gaps.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Survey and Assessment of Computational Capabilities for Advanced (Non-LWR) Reactor Mechanistic Source Term Analysis.

A vital part of the licensing process for advanced (non-LWR) nuclear reactor developers in the United States is the assessment of the reactor’s source term, i.e., the potential release of radionuclides from the reactor system to the environment during normal operations and accident sequences. In comparison to source term assessments which follow a bounding approach with conservative assumptions, a mechanistic approach to modeling radionuclide transport, which realistically accounts for transport and retention phenomena, is expected to be used for advanced reactor systems. As the designs of advanced reactors increase in maturity and progress towards licensing, there is a need to advance modeling and simulation capabilities in analyzing the mechanistic source term (MST) of a prospective reactor concept. In the present work, a survey is provided of existing computational capabilities for the modeling of advanced reactors MSTs. The following reactors are considered: high temperature gas reactors (HTGR); molten salt reactors (MSR) which include salt-fueled reactors and fluoride salt-cooled high temperature reactors (FHR); and sodium- and lead-cooled fast reactors (SFR, LFR). A review of relevant codes which may be useful in providing information to MST analyses is also completed, including codes that have been used for source term analyses of LWRs, as well as those being developed for other aspects of advanced reactor system modeling such as reactor physics, thermal hydraulics, and chemistry. A discussion of MST modeling capabilities for each reactor type is provided with additional focus on important phenomena and functional requirements. Additionally, a comprehensive survey is provided of tools for consequence modeling such as atmospheric transport and dispersion (ATD).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Mechanistic Source Term Considerations for Advanced Non-LWRs (Revision 1)

This report is a functional review of the radionuclide containment strategies of fluoride-salt-cooled high temperature reactor (FHR), molten salt reactor (MSR) and high temperature gas reactor (HTGR) systems. This analysis serves as a starting point for further, more in-depth analyses geared towards identifying phenomenological gaps that still exist, hindering the creation of a mechanistic source term for these reactor types. As background information to this review, an overview of how a mechanistic source term is created and used for consequence assessment necessary for licensing is provided. How a mechanistic source term is used within the Licensing Modernization Project (LMP) is also provided. Lastly, the characteristics of non-LWR mechanistic source terms are examined. This report does not assess the viability of any software system for use with advanced reactor designs, but instead covers system function requirements. Future work within the Nuclear Energy Advanced Modeling and Simulations (NEAMS) program will address such gaps. This document is an update of SAND 2020-6730. An additional chapter is included as well as edits to original content.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

AGR Equipment and Materials Disposition

Fabrication of low-enriched experimental nuclear reactor fuel for the Advanced Gas Reactor (AGR) Fuel Development and Qualification program has concluded after several years of effort. Residual chemicals, uranium compounds, fuel intermediates and fuel compacts need to be dispositioned in an economically and environmentally sound manner that maximizes, within reason, the preservation and repurposing of valuable materials. Government Furnished Equipment (GFE), procured with AGR program funds, are also to be dispositioned as economically as reasonably possible while seeking to maintain the functionality of BWXT equipment and repurposing of stand-alone GFE. Any stand-alone equipment that cannot be repurposed will either be abandoned in place, with BWXT consent, or discarded to an appropriate waste repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Demonstration of ACCERT Software for Nuclear Power Plant Techno-Economics

In the past few years, there has been a renewed interest in the deployment of nuclear power for decarbonizing the electricity grid as well as a range of industrial applications. As the demonstrations of advanced nuclear power plants start to begin, there will likely be a further increase in this interest. As nuclear is being considered as a part of the energy mix, understanding the cost of nuclear energy becomes increasingly important for all stakeholders including advanced reactor vendors (for making design decisions and marketing their designs), users of nuclear energy (e.g., to estimate the cost of decarbonization of other industries using nuclear), and government (e.g., in capacity expansion models that are used in framing policy). In this summary, we demonstrate a software tool called ACCERT that is currently being developed with funding from the Systems Analysis and Integration (SA&I) program under the Department of Energy’s Office of Nuclear Energy (DOE NE). ACCERT is a cost estimation and techno-economics tool for nuclear power plant applications that includes a database of (a) cost estimates of various ‘reference’ nuclear power plant designs gathered from existing literature, and (b) algorithms developed from these costs that can be used extrapolate the existing costs and perform a bottom-up cost estimation of other designs. A companion summary describes the software and its design in more detail and this summary presents a demonstration for four different nuclear power plant designs: a pressurized water reactor (PWR), high-temperature gas reactor (HTGR), sodium fast reactor (SFR), and a heat-pipe microreactor. The demonstrations include the reference cost estimates and the cost estimates of a modified design for each reference case.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Graphite for Advanced Nuclear Reactors: Deployment Readiness Review

Historically, graphite has been used in numerous reactor technologies, including research/test reactors and, commercially, in advanced gas reactors. These non-metallic materials can play a key role as internal core structures, reflectors, or neutron moderators, making them important for deployment of certain advanced reactor technologies. This report explores the industry readiness for graphite material deployment, documenting Codes and Standards applicable to their design, qualification, and manufacturing. Discussions also examine the manufacturing processes, aging/degradation, inspection techniques, and disposal options. The report covers the current status, documents gaps, and proposes some conclusions about approaches to managing some of the current gaps.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Comparative Study of Laboratory-Scale and Prototypic Production-Scale Fuel Fabrication Processes and Product Characteristics

An objective of the High Temperature Gas Reactor fuel development and qualification program for the United States Department of Energy has been to qualify fuel fabricated in prototypic production-scale equipment. The quality and characteristics of the tristructural isotropic coatings on fuel kernels are influenced by the equipment scale and processing parameters. Some characteristics affecting product quality were suppressed while others have become more significant in the larger equipment. Changes to the composition and method of producing resinated graphite matrix material has eliminated the use of hazardous, flammable liquids and enabled it to be procured as a vendor-supplied feed stock. A new method of overcoating TRISO particles with the resinated graphite matrix eliminates the use of hazardous, flammable liquids, produces highly spherical particles with a narrow size distribution, and attains product yields in excess of 99%. Compact fabrication processes have been scaled-up and automated with relatively minor changes to compact quality to manual laboratory-scale processes. The impact on statistical variability of the processes and the products as equipment was scaled are discussed. The prototypic production-scale processes produce test fuels that meet fuel quality specifications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

AGR 5/6/7 Data Qualification Report for ATR Cycles 162B through 168A

This report provides the qualification status of experimental data for the Advanced Gas Reactor (AGR) 5/6/7 fuel irradiation. AGR-5/6/7 was conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tri-structural isotropic (TRISO) low-enriched fuel for use in high temperature gas-cooled reactors. The objectives of the AGR-5/6/7 experiments are to: (i) irradiate reference-design fuel particles to support fuel qualification, (ii) establish operating margins for the fuel beyond normal operating conditions, and (iii) provide irradiated-fuel performance data and irradiated-fuel samples for post-irradiation examination (PIE) and safety testing. The test train contains five separate capsules that were independently controlled and monitored. Each capsule contains multiple 12.51-mm-long compacts filled with low enriched uranium carbide/oxide (UCO) TRISO fuel particles. The primary objective of the AGR-5/6 test (Capsules 1, 2, 4, and 5) is to verify successful performance of the reference-design fuel under normal operating conditions. The AGR-7 test (Capsule 3) was designed to explore fuel performance at higher temperatures to demonstrate the capability of the fuel to withstand conditions beyond normal operating conditions in support of plant design and licensing. AGR 5/6/7 will also provide irradiated-fuel performance data on fission-gas release from failed particles during irradiation. The AGR-5/6/7 capsules were irradiated in the ATR northeast flux trap location. The experiment began on February 16, 2018 and ended on July 22, 2020, spanning nine ATR cycles over two and a half years. Thus, the AGR-5/6/7 fuel compacts were irradiated for a total of 360.9 effective full power days. The AGR 5/6/7 experiment was able to remain in the reactor core during all three Powered Axial Locator Mechanism (PALM) cycles (163A, 165A, and 167A) without overheating its fuel compacts. This report includes irradiation monitoring data from nine ATR Cycles: 162B, 163A, 164A, 164B, 165A, 166A, 166B, 167A, and 168A, as stored in the Nuclear Data Management and Analysis System (NDMAS). During irradiation, data records consisted of instantaneous measurements recorded every minute and provided by text files automatically every 2 hours. The AGR 5/6/7 data streams addressed in this report include thermocouple (TC) temperatures, sweep gas data (flow rates [capsule inlet, outlet, and downstream at detector], pressure, and moisture content), and Fission Product Monitoring System (FPMS) data (release rates and release to birth rate ratios [R/Bs]) for each of the five capsules. A total of 94,989,908 TC temperature and sweep gas data records were received and processed by NDMAS for AGR 5/6/7 irradiation. Of these records, 41,593,387 (or 43.7% of the total) met data collection and accuracy requirements and are labeled as Qualified. A total of 57,746,693 TC temperature readings were captured from 54 installed TCs. Among them, 10,034,676 TC temperature records (only 17.4%) were Qualified and 47,701,371 TC temperatures (or 82.6%) are Failed due to 48 TC failures (63.5%) and due to missing values (19.1%). To assess performance of the operational TCs, analysis of daily correlations between TCs found no evidence of virtual junction failure for any TCs. Analyses on control charts of TC temperature differences revealed trending in TC readings for TC2, 4, 5, and 13 in Capsule 3, but there is no conclusive indication of TC drift failure that caused those trends. Therefore, TC control charts are not used to disqualify TC data, but only for users’ consideration. For sweep gas flow rates, a total of 31,519,747 gas flow records (84.4%) are Qualified for use for AGR-5/6/7 experiment; 5,723,468 gas flow records (15.4%) are Failed due mostly to missing values; and 74,641 high sweep gas flow rates (0.2 %) are Trend. A large number of Failed missing TC temperature and gas flow values were caused by an error in the data output script that outputted a ‘NULL’ value when values were unchanged. This problem was fixed during the outage of Cycle 166B, which led to a substantially decreased number of missing values during the last three cycles. Nonetheless, a large amount of non-missing data remained because of the high data acquisition frequency (1-minute) and still provided sufficient data to effectively monitor the experiment as designed. For FPMS data, NDMAS received and processed fission product release and R/B data for nine ATR cycles, when ATR core reached full power during AGR 5/6/7 irradiation. These data consist of 110,388 release rate records and 110,388 R/B records for the twelve radionuclides (Kr 85m, Kr 87, Kr 88, Kr 89, Kr 90, Xe 131m, Xe 133, Xe 135, Xe 135m, Xe 137, Xe 138, and Xe 139) for each of the five capsules. Equivalent numbers of uncertainty records associated the release rates and R/B values were provided. To date, qualification status of the FPMS data stored in the NDMAS dat

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Test Plan for the Third Reirradiation Heating Test of Fuel Particles from AGR 2 Compact 5-4-2

Post irradiation examination (PIE) and high-temperature heating tests of tristructural isotropic (TRISO)-coated fuel particles and compacts from the second Advanced Gas Reactor (AGR) irradiation experiment, AGR 2, are in progress. This test plan describes the reirradiation and subsequent high-temperature testing of loose fuel particles obtained from AGR 2 Compact 5-4-2. To simulate exposed kernels, four of these particles have been cracked at Oak Ridge National Laboratory (ORNL) to produce localized defects that extend through the TRISO layers (outer pyrolytic carbon, silicon carbide [SiC], and inner pyrolytic carbon) while keeping these layers in place and maintaining the shape and integrity of the kernel. To generate short-lived fission products, these particles will be reirradiated in the Neutron Radiography Reactor (NRAD) at the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL) Materials and Fuels Complex prior to heating testing in the Fuel Accident Condition Simulator (FACS) furnace at HFEF. FACS testing will be at 1200°C, and fission products released from the reirradiated particles will be measured. The intention of the test is to assess the retention of short-lived fission products I-131 (t1/2 = 8.02 days) and Xe-133 (t1/2 = 5.2 days) in the fuel kernels at high temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Semi-Annual Report for Modular Integrated Gas High Temperature Reactor Development during Performance Period April 2022 - September 2022

Horizontal Compact High Temperature Gas Reactor (HC-HTGR) is being designed by a multidisciplinary team of nuclear, mechanical, and structural engineers under the support of a DOE-NE Advanced Reactor Demonstration Program’s Advanced Reactor Concepts-20 (ARC-20) award. The objective of this ARC-20 project is to deliver a conceptual design for the proposed HC-HTGR in 3 years and support its commercialization as a safe, low-cost HTGR. Argonne National Laboratory (Argonne) is responsible for the design and analysis of the reactor cavity cooling system (RCCS) as a safety system for passive decay heat removal of the reactor concept. Additionally, Argonne is providing analysis of the primary coolant system to ensure temperatures within the core remain below safety margins during steady-state and potential accident scenarios. This second semi-annual report summarized the progress made at Argonne on the two tasks during the second half of FY22. As a part of the RCCS design task, a scoping calculation in estimating HVAC capability for the HC-HTGR reactor building was performed. A water panel modeling study was first performed with the test case, which confirms the capability of the RELAP5-3D modeling approach to explore various design options of the HC-HTGR RCCS under consideration. Then, a reference RELAP5-3D model for the unit geometry of the preliminary design of the HC-HTGR RCCS was developed. A preliminary performance analysis was conducted to evaluate the performances of a single-phase natural circulation and panel conduction in various operating conditions. For the primary coolant system analysis task, preliminary thermal hydraulic analysis of the HC-HTGR core design was performed with a high resolution 1D fluid-3D solid coupled model using the System Analysis Module (SAM) to assess the assembly coolant channel and bypass flow mass flow rate distribution. Some preliminary work on the development of a full core reduced order model was discussed following the assembly level model to predict the core wide coolant flow distribution and to model certain operational and accidental transients.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗