SEARCH · Engineering Papers
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.
Geology & Geophysics Seminar Series, College of Earth, Ocean, and Atmospheric Sciences, Oregon State University, Corvallis, OR, February 2, 2023
A numerical wave tank approach for computational fluid dynamics (CFD) modelling of an extreme irregular seastate is presented. The technique couples a potential flow solution with a CFD solver for more efficient numerical predictions. This method has recently become attractive both for the research community and the industry working with offshore structures. The model is used to determine the response of a submerged pressure differential wave energy converter (WEC) in a fully nonlinear irregular wave condition using the high fidelity CFD code, STAR-CCM+. Potential flow based numerical models are commonly used to predict motions and performance of wave energy converters. Wave kinematics can deviate from potential flow predictions for extreme wave conditions; the excitation loads on an absorber can also be increasingly influenced by viscous effects, not captured in engineering level models. In these extreme conditions, a Reynolds-averaged Navier-Stokes CFD model can better predict motions and loads for a WEC. Long time series with varying random seed numbers can be used to identify singular extreme wave events from a stochastic irregular sea state. This approach simulates a more realistic wave series for a given sea state than a regular wave or a focused wave. However, it is computationally infeasible to run these long time series for three-dimensional (3D) CFD simulations. In this work, two-dimensional (2D) CFD simulations with a long domain allow the full development of an extreme nonlinear wave condition. The results are used to identify extreme events from a 50-year storm condition for the PacWave site off the coast of Oregon. A relatively short time window including this extreme event is then mapped to a 3D simulation using a user defined wave methodology. Convergence studies for domain length, wave forcing lengths, and time before the extreme event were conducted.
This paper presents preliminary results from a subset of work carried out as part of a multinational research project entitled DErisking Exploration for multiple geothermal Plays in magmatic ENvironments (DEEPEN), supported by the U.S. Department of Energy (DOE) and Geothermica, a joint effort by EU member states and associated countries. The DEEPEN project will develop a customized approach to exploration for supercritical and superhot geothermal plays in magmatic systems, which will be applied to two demonstration sites. This paper summarizes field activities carried out at the U.S. demonstration site, Newberry Volcano in central Oregon. The objective of this work effort is to refine the subsurface model of Newberry Volcano, with special focus on deeper zones including the magmatic plumbing system and other key geologic elements. New data collection included gravity and wideband magnetotelluric (MT) surveys, as well as reinstallation of a seismic network. The National Renewable Energy Laboratory (NREL) and Enthalpion Energy LLC (Enthalpion) worked with the Deschutes National Forest Fort Rock District to use a low ground disturbance method of MT deployment to collect MT data inside the caldera and other restricted areas inside the National Volcanic Monument. This opened these areas to geophysical exploration for the first time in decades. Sites along and adjacent to the south rim of the volcano constituted the primary survey objectives. A team from Lawrence Berkeley National Laboratory (LBNL), the U.S. Geological Survey (USGS), and AltaRock also began the process of reinstalling the seismic network from the AltaRock enhanced geothermal system (EGS) demonstration in anticipation of further development activities at the site. The data ingestion, reduction, and analysis phase of the project is ongoing. We are currently processing the MT and gravity data and are developing a new, highly GPU-accelerated, 3D joint MT and gravity inversion to better localize the south rim/south flank conductive target and better understand its relationship to deep heat, fluid sources, and surface extrusive features. Joint inversions, which have not yet been undertaken at Newberry, will allow us to obtain constraints on the geologic model that cannot be determined from each method in isolation, improving our ability to image key geologic features at depth.
Oregon State University Materials Science Seminar, Corvallis, OR, June 1, 2023
Cape Thompson, Alaska, is located approximately 120 miles north of Kotzebue, Alaska, within the Ogotoruk Creek valley. The remote Alaskan location is also home to the US Atomic Energy Commission (AEC) Project Chariot study area. Project Chariot was part of the Plowshare Program, created in 1957 by the AEC to study peaceful uses for atomic energy. Project Chariot began in 1958 when a scientific field team chose Cape Thompson as a potential site to excavate a harbor using a series of nuclear explosions. Between 1959 and 1962, AEC was responsible for over 40 pretest bio-environmental and geotechnical studies of the Cape Thompson area. Strong public opposition to the tests developed, which caused the project to be cancelled with no nuclear explosions ever conducted at the site. In the early 1960's, several deep boreholes were drilled to obtain subsurface, temperature, and high-explosive testing data. Refrigerated diesel was used as a drilling fluid to keep the boreholes open so that thermistors could measure subsurface temperatures. The test boreholes were named Able, Baker, Charlie, Dog, and X-1. In the summer of 2014, the US Department of Energy, Office of Legacy Management (LM) completed the remediation of remnant diesel at these sites. Prior to the 2014 remediation activities, LM conducted two separate characterization field studies to determine the nature and extent of the contamination from the 1960's drilling activities. In 2013, LM developed a remediation plan that was submitted and approved by the Alaska Department of Environmental Conservation (ADEC). The 2014 Project Chariot remediation had some unique logistical challenges and historic issues with stakeholders to overcome. Many of the historic issues centered on previous AEC scientific actions and a general mistrust of the government by the local stakeholders. Prior to and during the 2014 remediation work, an outreach campaign was implemented to discuss prior activities at the site and include stakeholders from Point Hope, Alaska, in the remediation-planning process. The remoteness of the region also provided unique logistical challenges. Because of the short field season, LM initiated several subcontracts over a year in advance. The field camp, heavy equipment, and soil storage containers were shipped by barge to the site. Personnel, food, and small materials were flown in by bush plane from Kotzebue, Alaska. Previous site characterization had identified diesel-range organics (DROs) as the primary contaminant. The cleanup levels for the site were provided by the Alaska Administrative Code (AAC) for the Arctic Zone region. The AAC cleanup level for DROs on gravel pads is 500 mg/kg and 12,500 mg/kg tundra in the arctic-zone region. To help determine the extent and direct soil excavation in the field, photoionization detectors (PIDs) were used for field screening. While the PID results provided rough limits for excavation, laboratory confirmation sampling was required to provide actual excavation guidance. All excavation locations remained open until laboratory results met AAC criteria and ADEC approved closure. The laboratory sample protocol required the bush plane to pick up samples almost daily and transfer them to Anchorage for analysis. In order to meet holding time requirements, all laboratory samples were analyzed on an expedited basis. Analytical results and ADEC approval notifications were e-mailed back to the site via satellite connection. A total of 202 laboratory analyses were needed to complete the remediation effort. At each location (Able, Baker, Charlie, Dog, and X-1), the boreholes were plugged and abandoned in accordance with ADEC regulations. Contaminated soil was excavated, and the material was placed in super-sacks and hauled to connex boxes located near the sea shore. Approximately 786 tons of material was removed and shipped by barge to the Columbia Ridge Landfill in Arlington, Oregon, for disposal. The well casings were cut and sealed below ground surface. Each location was graded to its original contours and reclaimed with native vegetation, and a geodetic survey was conducted. Tours with leaders of the Village of Point Hope and the land owner (Arctic Slope Regional Corporation) were conducted prior to the final demobilization. All remediation equipment and camp facilities were removed at the completion of the project. (authors)
Remediation and management of contaminated facilities, soil, sediment and biota is an essential element of the DOE-EM mission for site completion, contaminant footprint reduction, and protection of human health and the environment. Protecting ecological resources is an integral part of protecting human health because people are part of the ecosystem and rely on ecological resources for food, fiber, clean air and water, recreational, aesthetic, and cultural values. Evaluation and management require understanding the past and current conditions, as well as protecting, preserving, and enhancing ecological resources. Understanding past and current resources includes evaluation of individual and population health within a local and regional context. The objective of this research is to describe a field method of evaluating ecological resources, based on work at the Hanford Site, which can be applied to sustainability of ecological resources as another mandate of DOE. This requires objective methods to characterize and evaluate resources spatially, within a context of local and regional importance. The DOE complex is large, diverse, and multi-state. The habitats and ecological resources are diverse, and the relative importance of ecological resources to the local ecosystem, region, and human communities varies. Large and diverse sites such as Hanford, Idaho National Laboratory, Los Alamos, hold resources of substantial importance to local and regional ecosystems, as well as to the ecology and cultural values of Tribes and other residents. However, other smaller sites (e.g. Oak Ridge, Brookhaven) also hold key ecological resources to local and regional ecosystems, cultures, and economics. Understanding the relative importance of ecological resources on each remediation site at different spatial scales, within and among DOE-EM sites is critical to management, stewardship, and ensuring sustainability for generations to come. Without a standardized methodology it is difficult to develop sound remediation and management plans to protect human health and the environment. The Ecological Evaluation of Resources described in this paper is part of the Consortium for Risk Evaluation with Stakeholder Participation's (CRESP) Risk Review Project on the Hanford Site. It is an independent assessment based of on-site field evaluations that encompasses evaluations of site resources in comparison to the Columbia Basin Ecosystem prior evaluations made by DOE, the States of Washington and Oregon, Nature Conservancy and Tribes. The on-site field evaluations are the focus of the current paper. The on-site field evaluation methodology was developed to update existing information on habitat values in light of changing landscape features (edge, patch size and shape, connectivity) and the current presence of invasive species. Both aspects change with time, especially given that remediation of remediation units changes the landscape. That is, every time there is a remediation and restoration of an operable unit (remediation site) on a DOE site, it changes the adjacent landscape (and ecological resources). The changes are greater where new roads were built to access a remediation site, staging areas constructed, and as a function of the size and time-frame for the remediation. Since ecological resources form a continuum across the landscape, and activities from remediation on a given site can affect those on adjacent areas, we evaluated resources both on the remediation site itself, and on a buffer equal to the largest diameter of the remediation site. The field evaluation methodology developed at Hanford Site makes use of existing resource level maps (DOE/RL-96-32 2013) and field surveys and measurements of current vegetation and habitat conditions to evaluate potential ecological impacts associated with cleanup activities at the evaluation unit. Additional information used in the ecological resource evaluations include current Endangered and Threatened Species distribution data (both state and federal), aerial imagery, locations of waste sites and infrastructure, and species and habitats of special concern. We suggest that the field methodology developed for the Hanford Site could be used across the DOE complex to evaluate ecological resources. This would provide DOE-EM with a common data base to make long-term decisions about the protection and sustainability of resources on the DOE complex, and would provide Tribes and the public with information on ecological resources on DOE sites. Such information could be provided in a web-based format which DOE, regulators, Tribes, and the public could easily access. (authors)
Currently, flow testing is being performed at Oregon State University (OSU) on the hydromechanical Fuel Test Facility (HMFTF) and the Endurance Flow Loop (EFL) to aid in the qualification of low enriched plate-type Uranium fuels for use in high performance research reactors in the US, under the Global Threat Reduction Initiative (GTRI). Under this project, varying geometries of small capsules proposed for irradiation testing in the Advanced Test Reactor (ATR) are being hydraulically testing prior to being placed in the ATR. The premise of this study was to conduct a number of flow tests on these capsules to characterize the pressure loss through each as a function of flow rate and compare the results against predicted theoretical values. Two categories of flow tests were performed in the Endurance Flow Loop to supplement flow testing of proposed LEU plate-type fuel for conversion of High Performance Research Reactors: Capsule Comparison Tests and Orifice Characterization Tests. The purpose of these tests were to determine the pressure drop across each experiment as a function of flow rate and to compare the results against a simple analytical model of the pressure loss through a pipe, to support the safety basis for the in-pile experiments. The capsule comparison tests compared the pressure loss through three types of capsules: RERTR, MP-1, and EMPIRE. In this category, nine flow tests were performed to determine the effect of having one, two, and three of each type of capsule placed in series in a test section. The analytical model accurately predicted, for each case, an increase in hydraulic resistance for each capsule that was added, though the values were conservative. The orifice characterization tests compared the pressure loss through various configurations of small orifice spacers in conjunction with blank spacers and the MP-1 capsules. The analytical model predicted that adding blank spacers would slightly increase the pressure drop due to added frictional effects, but indicated no change in the pressure drop based on the orientation of the capsules. The measured results indicated that the addition of the blank spacers had no measurable effect on the pressure drop, while changing the orientation of the capsules did. In all 15 flow tests, the theoretical model provides a conservative estimate of the pressure drop through each configuration for an equivalent flow rate, over that of the measured results.
There are a number of applications in nuclear energy and hazardous waste disposal that require monitoring of fluids under extreme environments, including high levels of temperature, pressure, toxicity and radioactivity. Many of these applications will benefit from a monitoring technique that is external and non-invasive. Currently the sensors used are invasive, must reside inside the pressurized vessels and must penetrate the vessel walls, which can create a weakness in the vessel. Additionally, instruments that are used inside such containers must be exceptionally hardened to the environment. Information Systems Laboratories (ISL) has developed an external mass (gravimetric) measuring technique for monitoring nuclear coolant in Small Modular Reactors (SMRs), which will also work for measuring fluid levels in waste tanks, that avoids the problems inherent in invasive sensors. It utilizes a COTS gravitational sensor of unprecedented accuracy, leveraged via proper sensor placement geometry, to detect fluid changes of small amplitude from an outside position, obviating the need to penetrate the vessel. The technique is called Gravisense{sup TM}. ISL has proven via simulation and experiment that this concept can be usefully applied to monitoring fluid levels in both nuclear reactors and large waste tanks. Numerical simulation algorithms were developed to calculate the gravity effect of small changes in water level, which were verified by experiments at the NIST Physical Simulator facility at the Oregon State University. The measured ultralow noise levels of the superconducting gravimeter type which utilizes a Niobium sphere suspended in a magnetic field to attain its phenomenal accuracy, demonstrated that fluid levels in SMRs can be measured at least to within 3 cm. Furthermore, the method can distinguish between a contained leak (from reactor to containment vessel) from an external leak (from reactor to outside of containment). Additionally, simulations of waste canisters that hold spent fuel rods show that the fluid level measuring accuracy can potentially do better than 1 cm accuracy by measuring from below the vessel, and judicious placement of sensors on top of large waste tanks can potentially achieve a very impressive 2 mm measurement accuracy. These encouraging results prove that the Gravisense{sup TM} technique for fluid determination can be very useful in nuclear energy generation, testing, and research, as well as in waste monitoring situations that are difficult to monitor via traditional sensing technology. We believe that the next step should be to test the technique on canisters of the type that are currently storing waste in various DOE locations. (authors)
The High Temperature Test Facility (HTTF) is a scaled, electric model at Oregon State University (OSU) of the Modular High Temperature Gas Reactor (MHTGR) by General Atomics. There are a few major differences between the design and the OSU facility. In the HTTF the reactor pressure vessel and the reactor cavity are in two separate tanks connected by a crossover duct similar to the crossover duct that would be seen between the reactor core and the steam generator in the MHTGR. This feature allows for the facility to open two main break valves simulating a Double-Ended Guillotine Break (DEGB) of the cross over duct, synonymous to a crossover duct break in the MHTGR. The density gradient driven flow phenomena in the HTTF duct can then be studied further. This break would cause a depressurization of the reactor and a loss of flow. It is called the Depressurized Loss of Forced Convection Accident (D-LOFC), also known as the Depressurized Conduction Cool-Down Event for High Temperature Gas Reactors (HTGRs) because heat removal is by conduction through the core and radiation out of the vessel. The HTTF was built and scaled to maintain the highest accuracy for the DCC. Prototypical temperatures with an inlet temperature of 259 deg. C and an outlet temperature of 687 deg. C are maintained in the HTTF. The temperature profile is also maintained in the smaller volume by using a ceramic core. The HTTF uses graphite heaters because the resistive heat produced does not change significantly with temperature . In order to produce a facility of similar temperatures within the allotted budget, sacrifices had to be made in other areas. The HTTF is scaled one-fourth in diameter, one-fourth in height and one-eighth in pressure of the MHTGR. There are also a few limitations for the HTTF. While in the MHTGR, the cavity is filled with air, the HTTF uses highly pure Nitrogen and does not have the ability to simulate a full depressurization. The DCC event is traditionally comprised of multiple stages: depressurization, molecular diffusion and natural circulation. However, this traditional interpretation ignores lock exchange flow which will occur immediately following depressurization and could have significant consequences in the reactor. It was found that the traditional model underestimates the amount of air ingress by at least 660 times according to Oh and Kim in their theoretical research of air ingress. Molecular diffusion also decreases with stratification. Any model that specifies the gases are stratified has further underestimated the air ingress of the DCC event. In addition, molecular diffusion is a much slower process, allowing the reactor to shut down and cool off considerably from operating temperatures. Thus, understanding lock exchange flow, the air ingress concentrations and the time scale is critical to MHTGR research and licensing. (authors)
Fuel supply stability is crucial to all reactors. The WWRSM research reactor operated by the Institute of Nuclear Physics at the Academy of Sciences in Uzbekistan currently uses Russian-supplied IRT-4M fuel. The goal of this Oregon State/PNNL research collaboration is to investigate the potential use of the Korean-made High-flux Advanced Neutron Application Reactor (HANARO) fuel rods in the WWR-SM. Many research reactors use unique fuels made by single suppliers, like the WWR-SM. The WWR-SM is a pool-type, high power research reactor using IRT-4M fuel assemblies which are sold and manufactured by a single fuel supplier. Use of HANARO fuel rods has the potential to improve fuel supply stability and associated costs while maintaining high flux levels needed for continuing operations. In the course of the investigation to design an alternate fuel assembly for the WWR-SM, multiple designs have been evaluated. In particular, two assemblies using standard HANARO fuel rods were considered: one with the optimal fuel-to-moderator ratio in the fueled region of the assembly and one with an under-moderated fuel region. Each of these assemblies has had its neutronic performance characterized with MCNP6 system reactivity and flux distribution calculations. (authors)
The next generation of nuclear power plants, such as Modular High Temperature Gas-Cooled Reactors (MHTGR), have passive safety design features to overcome transients during off-normal operation but it is unproven whether their safety systems will perform as per design requirements. The design and construction of a facility, to investigate the passive safety response of the complete reactor system will require significant resources. Therefore, a practical approach is to improve the understanding of this multi-physics or mixed effect problem in laboratory scale set-ups and computational models. The HTGR related separate effects experiments at academic institutions and other laboratories have been developed to provide validation data and complement computational efforts to identify critical issues related to passive safety design. There are several separate effect experimental studies currently under investigation or recently conducted by several academic institutions in U.S and throughout the world. The High Temperature Test Facility (HTTF) has been constructed at Oregon State University to conduct integral thermal-hydraulics experiments for the development of HTGRs. For Depressurized Conduction Cooling (DCC) experiments in HTTF, a mixture of nitrogen and helium gas is used to study internal natural convection up to 650?C. Similarly many other experimental studies have emphasized on HTGR core and plenum-to-plenum thermal-hydraulic behavior but with inert gases such as nitrogen or non-oxidizing solid substitutes for graphite. There is limited experimental data of mixed oxidation effects upon air ingress in nuclear grade graphite leading changes to thermal properties, heat generation due to chemical reactions and natural convection of air-helium mixture inside core at high temperatures (above 900?C). This work describes a unique experimental set-up developed by authors to conduct these mixed effect experiments. The current work is equally relevant to the designs well known by other names or geometries such as Very High Temperature Reactors (VHTRs) and Pebble Bed Modular Reactors (PBMRs). A common feature of HTGRs with passive safety design is that under certain accident conditions the decay heat is dissipated passively from the core by radiation/conduction/convection heat transport to the surrounding environment. Passive decay heat removal is one of the prime passive safety features of HTGRs and the heat conduction through the internal core region, graphite matrix, to the periphery is one of the major mechanisms of passive safety in HTGRs. The conduction and radiation heat transfer provide long-term heat removal but, during a Depressurized Loss of Forced Coolant (DLOFC) accident, the anticipated break in the coolant system will allow air ingress into the hot core region and can lead to graphite oxidation, which will negatively impact the core thermal behavior. In addition, air will also impact natural circulation of Helium and thus diminish the heat removal rate, which can further increase the graphite temperature. Therefore, both air ingress can eventually give rise to thermal instabilities and then to possible situations for breaching the safety envelope of TRISO fuel. This research summary is organized to present the background on oxidation studies, description of experimental setup and discussion of results in the following sections. (authors)
Currently, flow testing is being performed at Oregon State University (OSU) on the Hydro-Mechanical Fuel Test Facility (HMFTF) to aid in the qualification of low enriched plate-type Uranium fuels for use in high performance research reactors (HPRR) in the US, under the Global Threat Reduction Initiative (GTRI). Under this project, varying fuel geometries proposed for irradiation testing in the Advanced Test Reactor (ATR) are being hydraulically tested in the HMFTF prior to being placed in the ATR; these tests included the use of high-resolution tri-axis accelerometers affixed to the experiment in multiple locations. The premise of this study was to conduct a vibration analysis on the test designated as miniplate1 (MP-1) Large B experiment which is planned for irradiation testing in the ATR's Large B experiment location. (authors)
This study investigates the creation of a new set of scaling parameters for the Oregon State University Multi-Application Small Light Water Reactor (MASLWR) scaled thermal hydraulic test facility which modeled the NuScale power SMR design. As part of a study being undertaken by Idaho National Lab involving nuclear reactor load following characteristics, full power operations need to be simulated, and therefore properly scaled. Presented here is the scaling analysis and plans for RELAP5-3D simulation. (authors)
Understanding transient critical heat flux (CHF) resulting from a Reactivity Insertion Accident (RIA) in nuclear systems is an extremely undeveloped field. Though many experiments have been done for CHF, remarkably few have been done for rods at PWR operating temperature and pressure (300 deg. C, 2200 psia) with a short time constant ramp in power. This leads to many concerns regarding understanding how RIA scenarios develop and evolve in commercial nuclear reactors. With a limited understanding of the phenomenology governing transient CHF, there is little hope in being able to accurately model these scenarios using today's thermal hydraulic codes. The Accident Tolerant Fuels (ATF) Program established by the Department of Energy (DOE) is seeking out new fuel and cladding systems that can be used to reduce the probability and consequences of severe accidents. To better understand this phenomenon, the DOE is restarting the Transient Reactor Test (TREAT) Facility at the Idaho National Laboratory (INL). TREAT will be able to provide representative neutronic boundary conditions for testing coupled thermal hydraulics on the Transient Water Environment Recirculating Loop (TWERL) to study transient CHF with the best representation possible. To support this task, Oregon State University (OSU) is designing and building a representative flow loop that will be able to mimic the conditions of TWERL (sans the neutronic boundary conditions) to create a benchmark study for existing lumped parameter thermal-hydraulic codes such as RELAP5 and TRACE. This flow loop, the Transient Reactor Test Facility (TRTL), will support the ATF program by providing much needed NQA-1 data for transient CHF at operating temperature and pressure.
The high-temperature gas-cooled reactor (HTGR) is an advanced reactor concept that has received considerable attention over the last 60 years. HTGRs boast high coolant outlet temperatures, passive safety, and large margins to fuel failure. Recent years have seen increasing interest in HTGRs, whether large-scale HTGRs for electricity or process heat applications or as microreactors serving remote communities. HTGRs have considerable potential to fulfill these applications, but modeling and simulation tools typically used for reactor safety analysis have not been validated for HTGR modeling. The High-Temperature Test Facility (HTTF) was constructed at Oregon State University (OSU) to provide validation data for HTGR modeling. Recently, Idaho National Laboratory (INL), OSU, Argonne National Laboratory, Canadian Nuclear Laboratories, and the University of Tennessee Knoxville have collaborated to develop an HTGR thermal hydraulics benchmark based on HTTF experiments. This summary presents results from a portion of that benchmark using RELAP5-3D
This study investigated the fast-charging infrastructure needed by 2030 to enable seamless electrified road trips to and from national parks and monuments in seven western states: Washington, Oregon, Idaho, Wyoming, Utah, Nevada, and Arizona. It also estimated impacts to the electric grid. The research team investigated how on-route charging infrastructure projections change with different parameters or assumptions, as do related charging loads and grid impacts. NREL conducted the study in partnership with utility service provider PacifiCorp and Utah State University as part of the Western Smart Regional EV Adoption and Infrastructure at Scale project.
Pronghorn is a MOOSE based thermal-hydraulics code developed at Idaho National Laboratory (INL) for advanced nuclear reactor analysis. It has been previously applied to model pebble-bed high temperature reactors (HTRs), liquid-metal cooled reactors, and molten salt reactors, among others. This work leverages the coarse-mesh modeling capabilities in Pronghorn to model the Oregon State University (OSU)'s High Temperature Test Facility (HTTF). The HTTF is a 1:4 height scaled-down facility of General Atomics' Modular High Temperature Gas-cooled Reactor (MHTGR). The facility is primarily built to generate data for code and model validation, and does not precisely replicate MHTGR conditions. Nevertheless, it encompasses the main physics associated with MHTGR transients.
Prismatic gas-cooled reactors are a technologically mature reactor concept of interest for near- to mid-term deployment. To accelerate the deployment of these reactors, the DOE's Advanced Reactor Technologies Gas-Cooled Reactor campaign is spearheading a thermal hydraulics code validation benchmark based on the High Temperature Test Facility at Oregon State University. This presentation provides an introduction to that benchmark and an overview of some of the RELAP5-3D validation activities stemming from that benchmark.