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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 505 records · Page 28
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DOE OSTI2016

Performance tests of helium circulator under abnormal cooling conditions - 18523

Helium circulator is one of the key equipment of High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) and the unique rotating equipment of primary loop. It transports heat generated by fission reactions from reactor core to steam generator by forcing helium circulating in primary loop. As the study on the characteristics of helium circulator in HTGR was limited and the advanced designs integrated with many new technologies, a prototype of helium circulator was manufactured. It was supported by Active Magnetic Bearings (AMBs), and was assembled in a pressure vessel. The prototype was tested in Engineering Test Facility - Helium Circulator (ETF-HC), which is a full scale, rated flow rate and power, engineering test loop. Although performance tests under normal operation condition were successfully fulfilled, the complete features and performance still need further verification, especially for abnormal cooling conditions. Thus performance tests under abnormal cooling operation conditions were carried out. Tests such as temporary coolant loss of motor and AMBs cooler and abnormal high-temperature cooling water condition were fulfilled. All performance tests results showed that helium circulator met the entire requirement of HTR-PM plant under abnormal cooling conditions and validated its safety and reliability. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Effects of wall thickness on the flow instability of supercritical pressure water in a tube - 18568

The subcritical steam generators in the current high-temperature-reactor pebble-bed module, HTR-PM, in China can be replaced by supercritical steam generators to better coordinate the reactors and the supercritical steam turbines to improve thermal efficiencies with no phase change at supercritical pressures. This study numerically studied the heat transfer and flow instability of supercritical pressure water in a vertical tube with zero- and finite-thickness walls. The effect of the wall thickness on the heat transfer of supercritical pressure water at steady state was negligible. In the transient calculations, the fluid flow rate oscillated intensely beyond a certain heat flux with zero-thickness wall, while the flows in the tube with the finite-thickness walls acted totally different. The wall properties including the density and specific heat were varied to analyze why the oscillations were suppressed when the finite-thickness wall was considered in the calculation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Mechanistic source terms for the XE-100 reactor XSTERM Code Development Approach - 18587

X-Energy is developing the Xe-100 pebble bed modular High Temperature Gas Reactor (HTGR) as an advanced nuclear power source. To achieve its safety objectives, the Xe-100 reactor relies on materials inherent characteristics and passive safety features. Key features include the inherent high temperature characteristics of TRISO-coated fuel particles, graphite moderator, and helium coolant, along with passive heat removal capability of a low-power-density core. This paper will describe the approach adopted by X-Energy to develop and validate mechanistic source terms for the Xe-100. The Mechanistic Source Term (MST) approach for the Xe-100 reactor will build upon that of the prior DOE Next Generation Nuclear Program (NGNP) (Refs. 1, 2, 3 and 4). MST development is closely related to Fuel Qualification (FQ), Licensing Basis Event (LBE) selection and Probabilistic Risk Assessment (PRA). This paper provides a brief overview of the MST road map process. Software plays a crucial role in establishing and quantifying MSTs required for licensing the Xe-100. An insight into X-Energy s comprehensive, integrated software development program for calculating time-dependent radionuclide releases from the Xe-100 over a wide range of events is provided. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

German HTR phase out - experience and remaining challenges - 18659

For more than 50 years the pebble bed High Temperature Reactor (HTR) is under development in Germany. Forschungszentrum Juelich, Aachen University, but also industrial companies and other organizations are, respectively were involved in this task, which was for about two decades something like a national showcase. More than ten years before the German decision to shut down all nuclear power plants in Germany there was a strong decrease in the interest on HTR-related activities. Reasons might be the lack of competitiveness of HTRs for electricity production only in a well-developed grid, the bad performance of the German demonstration plant THTR and the rapid increase of the general German Angst. Most of the ongoing HTR activities will come to an end in a relatively short time. Last experimental projects cover the graphite dust behavior. The analytical work is focused on the further development of the HTR Code Package (HCP). In this paper, the German experience, remaining challenges and lessons learned will be addressed. Especially aspects such as fuel fabrication and qualification, the performance of AVR and THTR, the HTR-Modul concept, economic aspects of HTR, the competition and/or complementarities to LWR, the statement of inherent HTR safety and the development of large components for turbine, steam generator, blower and bearings will be presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Full Integration of the Fission Product Release Module STACY into the HTR Code Package (HCP) - 18661

The HTR Code Package is under development since 2010 at Forschungszentrum Juelich and at Aachen University. The HCP combines the capabilities of individual legacy codes developed during the German HTR program in a highly integrated manner and contains new features. Within the HCP, all aspects related to fission product release and the redistribution within the primary circuit are handled by the module STACY (Source Term Analysis Code system). STACY is the successor of the codes FRESCO-I/II (dealing with fission product release) and PANAMA (dealing with fuel performance issues). Compared to these predecessor codes, STACY can take amongst others a time-dependent nuclide inventory and a radial temperature profile within the fuel element into account. So far, this version of STACY performs an individual fission product release calculation for a representative number of tracer pebbles. While this version of STACY still relies on codes like VSOP and MGT-3D and contains features such as basic fuel management which are also contained in the before mentioned codes, STACY now will be coupled to forms an integral part of the HCP. Thus, fission product release calculations will be examined with one consistent reactor input model. This allows assessing the release of fission products already during the design phase, besides parameters like the maximum fuel temperature during a transient. Furthermore the total nuclide inventory of a fuel element is distributed among fuel kernels, coating layers, graphite grains and pores of the surrounding matrix graphite. Formerly, this total inventory calculated by the burnup code TNT was distributed according to the distribution of Uranium. As a new feature, STACY now keeps track of the distributions of nuclide inventories within the fuel element. Especially for nuclides mainly formed by activation and decay processes like Ag-110m, this approach is more accurate. Further, the knowledge about the actual nuclide inventories improves the accuracy of fuel performance calculations. In addition, the internal coupling allows a feedback from the fission product calculation to other code modules. Nuclides amounts being released can be considered in e.g. the burnup calculation while calculating the source term of for example tritium or within the calculation which takes care of the redistribution of species by the coolant. Within this paper, the application of the STACY to HTR cores with special emphasis on fission product release to reactor models such as the HTR-Modul and HTR-10 will be discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

V and V of the HTR Code Package (HCP) as a comprehensive HTR steady state and transient safety analysis framework - 18664

The HTR code package (HCP) allows for the simulation of several safety-related aspects of a High Temperature Reactor core in a highly integrated manner. HCP currently couples the thermo-fluid dynamics and time dependent neutronics code MGT, the spectrum code TRISHA, the burn up code TNT, the fuel management code SHUFLE and the fission product release code STACY. During its development, state-of-the-art programming techniques and standards are applied. The cross sections in HCP are generated by a 0D- or 1D solver featuring an innovative approach to treat the double heterogeneity of pebble fuel. Also major improvements have been made to optimize the nuclear data library based on ENDF/B-VII. Two advanced fuel management models are provided by the code module SHUFLE that go far beyond the capabilities of existing system codes like VSOP. The source term analysis code module STACY is coupled to HCP providing release rate calculations with a high spatial resolution making use of the nuclide densities provided by TNT. An outstanding new feature of HCP is the possibility to simulate long-term operation scenarios based on OTTO or MEDUL fuel shuffling schemes as well as selected transients in one integrated code package. Even alternating steady state/transient/steady state simulations are possible. So with HCP different fuel strategies and their influence on various kinds of accidents can be examined with one consistent reactor model. This paper provides an overview of the development status of the HCP and reports about selected benchmark results. It is demonstrated that the new system code HCP is capable to replace existing stand-alone codes like VSOP, TINTE/MGT, FRESCO or PANAMA while introducing new features, which so far to our knowledge were not available in the field of HTR safety research. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Feature selection based on PCA and the transient identification in nuclear power plants - 289

For the sake of enhancing safety and achieving more economic benefits, it is very important to timely and correctly identify the transients during the operation of nuclear power plants. There are thousands of monitoring signals in a nuclear power plant. It is difficult to complete the calculations timely if all the related signals are chosen for an algorithm of transient identification. Moreover, the irrelevant state variables may also do harm to the correctness of the identification process. Thus, one of the key tasks for transient identification is to figure out effective methods to extract the features from state variables. In this paper, we will address an efficient method for feature selection and transient identification based on principal component analysis (PCA). The effectiveness of the method are illustrated by the tests based on the data from the simulator designed for the High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM), which is under construction in China and the construction will be completed around 2017. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Enhanced electrical penetration components meet the high safety standards required for China's HTR-PM Shidaowan and generation mPower small modular reactors - 300

Enhanced glass-to-metal feedthroughs will be used to carry key instrumentation and control (I and C) signals in China's high-temperature, gas-cooled pebble-bed modular reactor (HTR-PM) at Shidaowan, and in Small Modular Reactor (SMR) prototyping. Chinergy Co., Ltd and Jiamusi Electric Machine Co., Ltd will use the feedthroughs, also called electrical penetration assemblies (EPAs), at the Shidaowan twin-reactor high temperature reactor (HTR). These feedthroughs are manufactured using specialized processes that make them far more robust than counterparts made with organic or ceramic materials. Improved EPAs manufactured by SCHOTT are the only feedthrough solution that can be used in the primary loop of an HTR because organic seals cannot withstand the high temperatures and pressures of this environment. Glass-to-metal seals (GTMS) are made using inorganic, non-aging materials which enable a virtually unlimited lifespan in high-temperature, high-radiation and high-pressure applications. This I and C technology package is being enhanced by the development and testing of new mineral-insulated (MI) cable connectors and EPAs made of crystallized glass-ceramic combined with superalloys. The EPAs have a maintenance-free lifespan of over 60 years, preventing any outage time related to these components and subsequently reducing costs associated with shutdowns. They are far more robust than organic seals and provide increased safety at a lower lifetime cost. Resistance against high pressure and high temperature are especially ideal qualities that make EPAs a valuable choice for environments with high standards for fire protection, such as nuclear plants, subsea oil drilling, or any other underwater application. Due to their significant offering of safety-relevant qualities, the EPAs have also been chosen by BWXT for implementation in its mPower SMR. Small modular reactor requirements are complex and very challenging but are met by the new EPA components, which will help form a strong I and C safety chain. Unlike ordinary compression glass-to-metal seals, GTMS with crystallized glass-ceramic have the ability to withstand temperatures of over 350 deg. C / 662 deg. F and operating pressures of 3500 bar. Furthermore, their lifetime is qualified for up to 60 years, including pressure and radiation requirements in accordance with the latest IEC/IEEE standards. Higher-specification EPAs and MI cable connectors have been designed into severe accident I and C instruments, including advanced hydrogen and oxygen sensors that exceed the challenging requirements of both HTR and SMR applications. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Neutronics Code Benchmark for Fluoride-Salt-Cooled Reactors

The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor design that aims to improve on the efficiency, cost, safety and non-proliferation security of nuclear energy, while exploiting existing technologies to minimize the design and licensing burden. FHRs share many design features with the high-temperature gas-cooled reactor (HTGR), namely using the same fuel form: a bed of graphite spheres, each with thousands of small coated fuel particles (TRISO) dispersed throughout it. However, rather than using a gas as a coolant, the FHR uses a molten salt which changes the behavior of the system; in particular, from a neutronics perspective, the coolant is no longer effectively transparent to neutrons, but rather contributes both to neutron moderation and absorption. It is difficult to perform exploratory reactor physics experiments for new reactor concepts, as the regulation and cost surrounding a fuel experiment is often highly restrictive and prohibitive. As such, researchers and design engineers must rely heavily on simulation capability to develop a working knowledge for how the reactor physics of a pre-prototypical reactor might behave. It is then essential to have confidence that simulation codes provide credible results. As there are no instances of FHRs deployed to date, nor are there any experimental data available on neutron behavior of pebble fuel immersed in salt, the first step is to perform a code-to-code verification of idealized scenarios that capture the physics involved with an FHR system. This summary describes the development of a code-to-code benchmark for FHR with the purpose of code verification and presents preliminary results. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Mixed effects associated with air ingress in graphite channels

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)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Post-Irradiation Examination for Conventional and Advanced Nuclear Fuel

During irradiation in a nuclear reactor, the fuel pellet experiences non-uniform fission rate and strong thermomechanical loads, which result in alterations and restructuring occurring non-homogeneously in the fuel volume. The irradiated fuel is characterized by a highly heterogeneous structure, with a correspondingly heterogeneous distribution of phases and properties. It is a big technical challenge for a hot cell facility to be able to cover effectively a broad variety of fuel types and compounds, characterized by different compositions, physico-chemical properties, geometries and configurations. The first challenge is to provide testing environments and procedures which do not alter or degrade the original irradiated fuel properties and are technically suitable for hot cell or shielded facility operation. For instance, non-oxide fuel compounds require protective atmosphere to avoid sample oxidation during storage, preparation and testing. Moreover, suitable adaptations of tools and measuring devices must be available to characterize fuel configurations which differ from the conventional cylindrical pellet-cladding geometries (e.g. spherical fuel elements for pebble bed HTR and TRISO coated fuel particles). Finally, adequate scientific methods of investigation must be available to measure properties and study the behaviour of highly heterogeneous irradiated fuel. In addition to basic techniques for non-destructive and destructive examination of nuclear fuel systems (e.g. covering fuel and cladding or fuel and coatings, or other configurations), 'in-depth' methods are applied to determine physical, thermomechanical and micro-analytical properties of specific phases/features present in irradiated fuel. This paper presents relevant trends and examples from fuel safety studies performed at JRC-Karlsruhe and in other laboratories which highlight the 'added value' obtainable by applying combined techniques, high resolution tools or 3D methods of analysis for the experimental 'multiscale' characterization of irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Progress in Development of Fluoride-salt-cooled High-Temperature Reactors (FHRs)

Fluoride-salt-cooled High-temperature Reactors (FHRs) are a new class of reactors that use graphite-matrix coated-particle fuel with a clean liquid salt coolant. The largest university FHR program is an Integrated Research Project (IRP) sponsored by the U.S. Department of Energy and others that is led by the Massachusetts Institute of Technology (MIT) and includes the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The year 2016 was the fifth year of this program. We report herein some of the major results of work done in 2016. The base-line FHR concept studied in this IRP uses (1) graphite-matrix coated-particle fuel with a pebble bed geometry and (2) a clean liquid salt coolant-flibe (a mixture of {sup 7}LiF and BeF{sub 2}). The combination of a very high-temperature fuel with failure temperatures above 1650 deg. C, a coolant boiling point above 1400 deg. C, and a coolant that traps most fission products implies the capability to assure no major fuel failure under severe accident conditions. There are a family of FHR concepts using different fuel geometries and different fluoride salts. The power conversion system is a Nuclear Air-Brayton Combined Cycle (NACC) with peak power provided by natural gas or a Firebrick Resistance Heated Energy Storage (FIRES) system. This enables a base-load reactor to provide variable electricity to the grid that substantially increases revenue relative to base-load nuclear reactors. Most of the results reported herein are applicable to most FHR concepts as well as liquid-fueled molten salt reactors (MSRs)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Preliminary Development of a Nuclide Inventory Code for HTGR

Nuclide inventory calculation for a reactor core is the basis of radiation protection, shielding design and radioactive environmental impact assessment of the power plant. HTR-PM (high-temperature gas-cooled reactor pebble-bed module) under construction in China is one of the candidates for Gen-IV advanced nuclear power system.1 In the design stage of HTR-PM, KORIGEN has been utilized in the inventory calculation.2 However, the current version of KORIGEN was developed several decades ago. In order to improve the design capability of the analysis code package for HTR-PM, a new computer code for nuclide inventory calculation called NUIT (Nuclide Inventory Tool) is being developed based on the DEPTH3 code. The NUIT code tries to utilize accurate and efficient burnup solvers, incorporate well-validated and most recent nuclear data libraries, and provide user-friendly IO interface. This paper presents the preliminary work in the development of NUIT.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Improved Heat Transfer and Volume Scaling through Novel Heater Design

In 2014, the University of California, Berkeley, Nuclear Engineering Thermal Hydraulics Laboratory completed construction of the Compact Integral Effects Test (CIET) facility, a scaled thermal hydraulics scaled integral effects test facility designed to model the Mark-1 pebble-bed fluoride-salt-cooled high-temperature reactor (Mk1 PB-FHR or simply Mk1). CIET uses an organic simulant fluid, Dowtherm A, to match the Prandtl, Reynolds, Froude, and Grashof numbers of important fluoride salt candidates for molten salt reactors, such as flibe. The FHR reactor core is simulated using a resistively heated tube, which heats the oil flowing through the annulus between the heated tube and a non-heated inner tube. However, the volume fraction of the coolant in the CIET heater was dramatically lower than the volume fraction of the primary coolant in the Mk1 PB-FHR design (3% in CIET vs. 17% in Mk1) because the CIET heater was initially scaled to match an earlier FHR design. Also, the maximum core heater power in CIET is limited by the melting point of Teflon components in CIET, approximately 330 deg. C. For conservatism, the CIET heater power is limited to keep the maximum surface temperature of the heater below 250 deg. C. At the nominal mass flow rate of 0.18 kg/s, this power was about 5 kW. A higher heat transfer coefficient from the heater wall to the oil was needed to increase the effective power delivered to the oil. In order to alleviate these two limitations, a new heater insert was designed and constructed for CIET. This new heater insert uses a twisted tape to augment heat transfer. The twisted tape is supported inside a perforated inner tube that has the same diameter as the original inner tube of the heater, in order to make it compatible with CIET's original heater design. The heat transfer and pressure drop properties of the new heater insert designs were tested in CIET and compared to the previous design in terms of pressure drop and convective heat transfer coefficient. The new heater inserts are intended to increase the range of power levels and mass flow rates that can be tested in CIET in order to augment understanding of FHR thermal hydraulic phenomena. Also, obtaining different data points for CIET response dynamics under different heater coolant residence times will be valuable in evaluating the Mk1 design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Initial development of a generic fluoride salt-cooled reactor model

Fluoride high-temperature reactors (FHRs) are high-temperature, low-pressure reactor concepts that use tri-structural isotropic (TRISO) fuel and molten fluoride salt coolant. These reactors have the potential to provide both electrical power and high-temperature process heat. We used generic FHR parameters for a pebble-bed FHR to develop an initial model with fresh fuel for a generic FHR (gFHR) in MELCOR and SCALE (NEWT and KENO). In this paper, we present the development of our gFHR models, which will serve as the baseline for a sensitivity and uncertainty analysis to quantify the range of possible source terms for FHRs in severe accidents. We present MELCOR results for fuel and coolant temperatures through the core, a nodalization study for the steady-state thermal hydraulic model, and development of reactor physics models in SCALE. As this work progresses, these models will be used to calculate source terms for a loss-of-forced-flow accident and to conduct a sensitivity study on this accident to establish a range of possible source terms. SCALE will provide reactor physics parameters like isotopic inventory, decay heat generation, and temperature coefficients of reactivity. Using the uncertainty quantification tools within SCALE, we will generate distributions for those parameters and will use the uncertainty quantification code RAVEN or DAKOTA to sample those distributions in MELCOR to quantify the impact of reactor physics and thermal hydraulic uncertainties on FHR source terms. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Recent Advanced Reactor Multiphysics Model Highlights in the Virtual Test Bed (VTB)

The Virtual Test Bed (VTB) host over 30 distinct simulations that showcase state-of-the art capabilities across the national lab complex. An update on the status of models on the VTB is summarized here, along with a more detailed overview of select recent new capabilities to showcase. All of the major advanced reactor types are represented in the VTB. The first example consists of a multiphysics simulation to track the transport of species in Molten Salt Reactors using depletion, advection, and thermochemical calculations. The second consists of a coupled neutronic and thermal hydraulic simulation to validate a gas cooled reactor. The third consist of pebble-bed equilibrium model for a fluoride high-temperature reactor. The fourth is a high-fidelity neutronic and thermal hydraulic model of a liquid metal reactor assembly. And lastly the fifth consists of transient multiphysics simulations of heat pipe microreactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Modeling of a 1/20th Scaled-down Gas Reactor Using MOOSE-based Application Pronghorn

University of Idaho (UI) designed and built a 1/20th scaled-down HTGR system to examine the mixing and venting of helium and reactor cavity air after a loss-of-coolant, also known as LOCA, due to a break in the reactor pressure vessel. Computational fluid dynamics (CFD) models of the system were developed, but their computational cost is high for modeling long transient accident scenarios. The objective of the project is to develop a simulation of the scaled-down system using the MOOSE-based application Pronghorn. A numerical approach is needed to have a better understanding of the physics phenomena that takes place following a LOCA event, and it can guide plant designers to design parameters for the development of mitigation techniques. This scenario is applicable to all HTR designs including the pebble bed reactor concept.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Coupled Multiphysics Primary Loop Simulations of the Mk1-FHR in the Virtual Test Bed

To support advanced reactor demonstrations, the Virtual Test Bed (VTB) [1] repository hosts a wide range of challenge problems for showcasing modeling and simulation capabil- ities in support of advanced reactor demonstrations. This document presents a coupled multiphysics model of the Mark 1 pebble-bed fluoride-salt-cooled high-temperature reactor (PB-FHR). The analysis leverages NEAMS tools (Griffin [2], SAM [3], Pronghorn [4], and the MOOSE [5] heat con- duction module) for core neutronics, thermal hydraulics of the core and primary loop, and multiscale fuel performance simulations. The analysis was entirely created by coupling standalone simulations of the reactor that were previously available on the VTB. All input files and documentation de- veloped for this example are available on the VTB website: mooseframework.inl.gov/ virtual_test_bed/ . This model was featured in the National Reactor Innovation Cen- ter Tech Talk presented in December 2021.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗