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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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Analysis of Phenix end-of-life natural convection test with the MARS-LMR code

The end-of-life test of Phenix reactor performed by the CEA provided an opportunity to have reliable and valuable test data for the validation and verification of a SFR system analysis code. KAERI joined this international program for the analysis of Phenix end-of-life natural circulation test coordinated by the IAEA from 2008. The main objectives of this study were to evaluate the capability of existing SFR system analysis code MARS-LMR and to identify any limitation of the code. The analysis was performed in three stages: pre-test analysis, blind posttest analysis, and final post-test analysis. In the pre-test analysis, the design conditions provided by the CEA were used to obtain a prediction of the test. The blind post-test analysis was based on the test conditions measured during the tests but the test results were not provided from the CEA. The final post-test analysis was performed to predict the test results as accurate as possible by improving the previous modeling of the test. Based on the pre-test analysis and blind test analysis, the modeling for heat structures in the hot pool and cold pool, steel structures in the core, heat loss from roof and vessel, and the flow path at core outlet were reinforced in the final analysis. The results of the final post-test analysis could be characterized into three different phases. In the early phase, the MARS-LMR simulated the heat-up process correctly due to the enhanced heat structure modeling. In the mid phase before the opening of SG casing, the code reproduced the decrease of core outlet temperature successfully. Finally, in the later phase the increase of heat removal by the opening of the SG opening was well predicted with the MARS-LMR code. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Casting evaluation of U-Zr alloy system fuel slug for SFR prepared by injection casting method

Metal fuel slugs of U-Pu-Zr alloys for Sodium-cooled Fast Reactor (SFR) have conventionally been fabricated by a vacuum injection casting method. Recently, management of minor actinides (MA) became an important issue because direct disposal of the long-lived MA can be a long-term burden for a tentative repository up to several hundreds of thousand years. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long-lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. In order to prevent the evaporation of volatile elements such as Am, alternative fabrication methods of metal fuel slugs have been studied applying gravity casting, and improved injection casting in KAERI, including melting under inert atmosphere. And then, metal fuel slugs were examined with casting soundness, density, chemical analysis, particle size distribution and microstructural characteristics. Based on these results there is a high level of confidence that Am losses will also be effectively controlled by application of a modest amount of overpressure. A surrogate fuel slug was generally soundly cast by improved injection casting method, melted fuel material under inert atmosphere.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Accident tolerant, non-oxidizable MOX fuel fabrication

Thorium dioxide (ThO{sub 2}) has a set of physical-chemical properties that make it well suited to serve as a nuclear fuel matrix: it has a very high melting point, it cannot be further oxidized, it has low leachability in water, reasonable thermal conductivity (better than UO{sub 2}), and of course it is fertile. Collectively, these properties allow for a fuel ceramic with increased in / out of core stability. As such, Thor Energy and a consortium of entities that include Westinghouse, KAERI, Fortum, NNL, ITU and IFE have recently fabricated (Th,Pu)O{sub 2} pellet specimens in Norway. The Th-MOX production methodology was developed over the course of 2 years of ceramic production trials involving Ce-Th oxide mixtures. It was understood that parameters developed for CeO{sub 2} material might need to have been modified if trial results with PuO{sub 2} led to sub-optimal pellets. Among observations made in this work, the most important production parameter was the preparation of constituent / mixed powders. Correct powder morphology was key to achieving high fuel densities. The process parameters developed for (Th,Ce)O{sub 2} pellets were adopted with the expectation that these would serve as a sound basis for obtaining dense, homogeneous (Th,Pu)O{sub 2} fuel ceramic. The Th-MOX fuel manufacture campaign was successful in producing enough (Th,Pu)O{sub 2} pellets to assemble two 30 cm long fuel pins for testing in the Halden test reactor. These fuel rods will be fitted with various instruments to enable temperature, fission gas release and dimensional changes to be measured while operating at power in conditions similar to those of a BWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

New Approach in Fissile Assay Using LSDS - 17076

A quantitative assay of isotopic fissile materials (U{sup 235}, Pu{sup 239}, Pu{sup 241}) was conducted at the Korea Atomic Energy Research Institute (KAERI), using a lead slowing down spectrometer (LSDS). Using a uranium rod and plutonium, the actual measurement of induced fission neutrons was performed in the RPI (Rensselaer Polytechnic Institute) LSDS facility. In the simulation, a content of U{sup 235}, Pu{sup 239}, and Pu{sup 241} was changed. The detected signal has a direct relationship with the mass of the existing fissile isotopes. The mathematical model for fissile assay was applied to the measured and simulated detection data. The assay results show that the isotopic content was consistent with 2 to 3% uncertainty for Pu{sup 239}. The simulation indicated that the assay result was affected by the self-shielding correction. Usually, in highly enriched plutonium, the correction plays an important role and good understanding of it is required in order to increase assay accuracy. The combination of LSDS technique measurement and simulation is a very powerful and direct way to analyze the isotopic fissile content. The information provided by such analysis is applicable to nuclear fuel cycle development and spent fuel management for safety and economics. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

PGSFR Core Design and Performance Characteristics

Korea Atomic Energy Research Institute (KAERI) has developed a Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) as a representative of Gen-IV concepts. The final PGSFR core will be composed of transuranic (TRU) fuels to transmute LWR spent fuels and self-recycled its fuels, in combination with a pyro-processing facility. However, the initial POSFR core consists of low enriched uranium (LEU) fuels in order to: 1) provide an irradiation test for TRU fuels and structural materials, 2) keep pace with the development of pyro-processing technology, and 3) obtain operational experience on the reactor systems and components. During the first phase of PGSFR development plan (2012-2015), the power level was decided to be 150 MWe from many trade-off studies using different power level. The PGSFR uranium core performance characteristics at the initial, transition and equilibrium states have been carried out. In this paper, the PGSFR core design and related performance characteristics are introduced. The PGSFR uranium core design was performed at the initial/transition/equilibrium core based on the candidate fuel management scheme. It was confirmed that this design can satisfy surely all design limit and requirement. The chapter 4 of PGSFR Preliminary Safety Information Document (PSID) published in December 2015 contains the detailed information and characteristics of PGSFR core. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Safety Analyses of Design Basis and Design Extended Condition Events for the PGSFR

KAERI (Korea Atomic Energy Research Institute) has been developing the PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor), which is a pool type sodium-cooled fast reactor with the thermal power of 392.2 MWt, uses metallic fuel of U-10%Zr for a core. The preliminary specific design of PGSFR is completed in 2015. The main parts of PGSFR consist of PHTS (Primary Heat Transport System), IHTS (Intermediate Heat Transport System), two SG (Steam Generator), and DHRS (Decay Heat Removal System). The main feature of PGSFR in safety is from the adoption of metallic fuel core, leading to inherent reactivity feedback mechanisms and high thermal conductivity. In this work, safety analysis on DBE (Design Basis Event) and DEC (Design Extended Condition event) for PGSFR preliminary specific design is conducted using MARS-LMR code. Conservative and best-estimated approaches are applied to DBA and DEC, respectively. It is found that the most severe accidents in the AOO, DBA class 1, DBA class 2, and DEC are LOF, OPS, TOP and ULOF event, respectively. And all analysis results are found to meet safety acceptance criteria requirements. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Convective Heat Transfer in Decay Heat Exchanger of Sodium-cooled Fast Reactor

Sodium-to-sodium heat exchangers such as the decay heat exchanger (DHX) are key components of the prototype generation [V sodium-cooled fast reactor (PGSFR). The DHX transfers the decay heat generated in the reactor core to the decay heat removal loop after a reactor shutdown. Shell-and-tube type with straight tube arrangement has been widely adopted for the sodium-to-sodium heat exchanger of the sodium-cooled fast reactor. Thus liquid metal heat transfer characteristics inside the tube and around the tube bundle have to be fully understood to design the sodium-to-sodium heat exchanger and evaluate its thermal performance accurately. Plenty of studies on the convective heat transfer of the liquid metals were conducted to present the Nusselt number (Nu) correlations with respect to the Peclet number (Pe). However, those correlations still deviated from each other considerably and few of correlations for low Pe range (less than 100) have been reported. Accordingly, it is difficult for a design code developer to choose a proper heat transfer coefficient correlation for the sodium-to-sodium heat exchanger particularly in a low flow rate range. Recently, a separate effect test for the DHX has been completed using the STELLA-1 facility at KAERI according to the STELLA program which aims at establishment of separate and integral test database (DB) for the design of the PGSFR. The heat transfer performance test of the DHX was successfully conducted under major operating conditions. The objective of this work is to suggest proper convective heat transfer correlations for tube- and shell-sides of the sodium-to-sodium heat exchanger based on the test DB obtained from STELLA-1. The Pe range of the analyzed experimental data is concentrated on less then 100. The adequacy of the suggested correlations is evaluated by applying to the DHX design code, SHXSA. Based on the DHX test DB of the STELLA-1, the proper convective heat transfer correlations for the tube- and shell-side of sodium-to-sodium heat exchangers were suggested. Two correlation combinations for each side were proposed and turned out to be reasonable through the evaluation using the SHXSA design code: Suggestion no. 1: - Lyon correlation (constant = 4) for tube-side; - Lubarsky-Kaufman correlation for shell-side. Suggestion no. 2: - Modified Lyon correlation (constant = 1.45) for tube-side; - Schad-modified correlation for shell-side The Pe ranges of the DHX test DB were 53 - 225 for the tube-side and 45 - 148 for the shell-side, but most of test conditions were concentrated on the Pe range of less than 100. Two correlation suggestions can be practically used for the design of sodium-to-sodium heat exchangers. However, by utilizing more test DB, it needs to be verified which one is closer to real phenomena. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Surface-modified microchannel-type process heat exchanger (PHE) for hydrogen production using sulfur-iodine cycle - 18481

For massive production of hydrogen using a very high temperature gas-cooled reactor (VHTR) system, a small-scale process heat exchanger (PHE) for feasibility testing of a sulfur-iodine (S-I) cycle is designed at KAERI. Printed circuit type heat exchanger (PCHE), which is suitable for a helium/decomposed sulfuric acid (SO{sub 3}/SO{sub 2}) heat transfer system, is employed with surface modification to protect the substrate alloy from the corrosive sulfuric acid environment. Experiments on diffusion bonding, which is a key technology in compact-type heat exchangers, are conducted to achieve considerable strength and ductility at both room and high temperature. Metallographic examination reveals the migrated grain boundary across the interface. Tensile ductility over 35 % is achieved and dimples are observed on the fracture surface in all testing temperatures. For surface modification, β-SiC having high corrosion resistance is deposited on the surface of the microchannels using electron beam physical vapor deposition (EB-PVD) followed by ion-beam irradiation to enhance the adherence at the interface of β-SiC/alloy. Through corrosion testing in aqueous sulfuric acid (60 vol.%), the weight loss of the surface-modified specimens is decreased by one order of magnitude, implying the relevance of coating for this application. (authors)

08 HYDROGEN↗
DOE OSTI2016

Preliminary Development of a Drift-Flux Model Based Pin-Resolved Core Thermal-Hydraulic Code for Dedicated Use in Efficient High-Fidelity Neutronics Coupled Calculations

As the high-fidelity multiphysics coupled core calculation becomes more practical with the advent of high-performance computing technology, efficient coupling of a neutronics code with a thermal-hydraulic (T/H) code at the pin level poses a new challenge. The COBRA-TF (CTF) code which is one of the well-known subchannel analysis codes based on the two-fluid, three-field model was coupled with many neutron transport codes such as MPACT, TORTTD, and nTRACER due to the easiness of coupling by its well-established coupling interface. Nevertheless, the calculation efficiency of such coupled codes is not satisfactory because of the heavy computational burden of CTF. The MATRA subchannel analysis code of KAERI which is based on the homogeneous equilibrium model (HEM) was coupled with neutronics codes as well. The choice of the HEM and the marching scheme in MATRA reduces the computational burden significantly. However, the applicability of the HEM for the two-phase flow conditions is limited to normal operating conditions of pressurized water reactors (PWRs) and therefore a better model which would allow wider applicability than HEM is necessary in order to deal with the core states involving nontrivial voiding. The drift-flux model (DFM) which uses 5 equations for two-fluid analyses by adopting the mixture velocity instead of two phasic velocities can be a good compromise between the two-fluid model and the HEM. With the DFM, the computational cost can be reduced and moreover the modeling of the interfacial area constitutive relations can be avoided. The phase separation for the mixture velocity can be incorporated by using the drift-flux parameters, C0 and Vgj, which can be obtained by using empirical formulas. Several subchannel codes such as WOSUB and ASSERT were developed based on the DFM. It was also adopted in some system analysis codes such as RETRAN3D. In the work here, the development of a DFM based pin-resolved T/H analysis code, ESCOT (Efficient Simulator of Core Thermal-hydraulics), is introduced for the use in efficient coupled calculations with a high-fidelity neutronics code. The field equations and the numerical methods are briefly described. Then the efficiency tests of various linear system solvers including the Krylov subspace methods (KSMs) are presented. A test problem is solved by ESCOT, MATRA and CTF for performance comparison and validation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Experimental Study of Molten Metallic Fuel Relocation in Sodium-Filled Fuel Structures

The Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) is being developed by Korea Atomic Energy Research Institute (KAERI). While their design incorporates inherent safety features demonstrated in the landmark Shutdown Heat Removal Test (SHRT) conducted at EBR-II including a large pool configuration and uranium alloy fuel, it is still necessary to evaluate metal fuel relocation and theirs consequences in case of postulated core disruptive accidents. Extensive data regarding fuel relocation in sodium from both in-pile and out-of-pile tests were obtained for oxide fuels in the 1960's - 1970's, while few studies of metallic fuel behavior were performed because of concerns that metallic fuel could not achieve high burnup due to the irradiation induced swelling problem. Following new metal fuel design development that mitigated the swelling problem, experiments regarding interaction of molten metal fuels and liquid sodium have been conducted by several researchers. Most studies that were conducted using metal simulants in various liquids concluded that the molten metals were heavily fragmented after liquid contact and the dominant factors governing the fragmentation were superheating and latent heat of fusion of the molten metal fuel. Gabor et al. conducted experiments using kilogram quantities of various molten uranium alloys in an open sodium pool configuration. They found that the metal fuel fragments were basically in forms of filaments and sheets with a high bed voidage in the order of 0.9. Their calculation indicated that the debris beds formed by relocated metal fuels in an open sodium pool would be largely coolable by conduction heat transfer; and even if deep beds were to form, convective heat transfer and boiling heat transfer could preclude further melt penetration. Although Gabor's tests are useful in evaluating the coolability of relocated metal fuels during a postulated accident in a SFR, they focused on late phase behavior in which the core material relocates downward through the core support plate into the inlet plenum where the structures are more open and thus consistent with the open pool configuration. There is still a knowledge gap related to nature of metal fuel fragmentation and relocation within the fuel assembly structures. Argonne National Laboratory (ANL) is planning to conduct experiments regarding metal fuel relocation in the core structures which mock up the PGSFR's fuel assembly design. In the present study, preliminary tests for demonstration of melting and injecting the metal fuel were conducted using copper. Metal fuel relocation experiments with single fuel pin test section are planned to be conducted for understanding the flow and freezing behavior of the molten metal fuel in a narrow sodium channel and for evaluating the coolability of the relocated fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Analysis on the Heat Transfer Phenomena Inside the Scaled-Down VHTR RCCS Riser Experiment

Very High Temperature gas-cooled Reactor (VHTR) that is one of GEN-IV reactors has been developed for the purpose of hydrogen production. VHTR is characterized by high operating temperature and stepped-up inherent safety given by passive safety systems such as Reactor Cavity Cooling System (RCCS). In case the active heat removal system losses its capability, RCCS significantly contributes to the inherent safety of VHTR by removing the decay heat transferred from the reactor core to the reactor vessel. Therefore, the performance of RCCS is the key parameter to ensure the safety of the reactor. Several computation and experiment researches on the performance of RCCS have been conducted. There are two types of RCCS depending on its coolant; one is the air-cooled type and the other is water-cooled type. This study focused on the former type. Due to the excessive height of the prototype RCCS, experiment facilities were constructed in reduced scale. Korea Atomic Energy Research Institute (KAERI) constructed the quarter-scale RCCS experiment facility, NACEF and Argonne National Laboratory (ANL) has constructed half-scale experiment facility, NSTF. University of Wisconsin (UW) also has the quarter-scale experiment facility and Texas A and M University has conducted separate effect test on the jet effect at the upper plenum of RCCS. From the reduced scale experiment result, the performance of RCCS such as a heat removal rate of RCCS and a peak temperature of reactor vessel were evaluated. When the reduced-scale result is applied to the prototype, however, a distortion could be brought about due to the reduced height of RCCS in the experiments. Especially, if the heat transfer regime inside the riser which roles as a heat exchanger to discharge the decay heat into the atmosphere is not identical, the riser wall temperature and the consequent reactor vessel temperature of the prototype could be incorrectly predicted. Thus, the convective heat transfer phenomena inside the riser duct both in the reduced scale and in the prototype is necessary to be investigated. In this study, the preliminary analysis on the heat transfer phenomena inside the RCCS riser was conducted. The subjects of the analysis were the prototype RCCS riser of PMR 200, and its scaled-down experiment facilities, NACEF. First, the geometry, flow and thermal conditions of the riser in PMR200 RCCS and NACEF were introduced with a scaling method. And then, the heat transfer regimes and the heat transfer coefficient corresponding to each condition were predicted using correlation and CFD calculation. Finally, planned activities including the experiment was introduced. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Chemical Decontamination Performance of Radioactive Metal Surface Using HyBRID Process

The metal oxides formed on the surface of the primary coolant system of a NPP (Nuclear Power Plant) contain radioactive components, and thus the oxides are generally need to be removed by means of chemical decontamination to reduce the exposure of workers during the decommissioning of a NPP. The commercial application of chemical decontamination technology has expanded significantly in recent years at NPPs. From initial applications to components and small subsystems, to the current examination of full system decontamination, the entire technology has become sophisticated. Dilute chemical decontamination solutions have been developed because too aggressive solution is deemed to cause the undesirable corrosion damage to the base metal as well as generate a large volume of decontamination waste. KAERI (Korea Atomic Energy Research Institute) has been developing a new multi-step chemical decontamination process for the primary coolant system of a NPP. This process does not contain any organic chemicals in decontamination solution. The HyBRID (Hydrazine Based Reductive metal Ion Decontamination) process which contains hydrazine as a strong reductant and cupric ion as a catalyst has been developed to be applied as a reductive decontamination step. Instead of usual NP (Nitric acid Permanganate) process, slightly altered SP (Sulfuric acid Permanganate) process is used as a pre-oxidative decontamination step in conjunction with HyBRID process. The preliminary results have proven the effectiveness of this process for the system decontamination. The present study is concerned with the verification of chemical decontamination efficiency for radioactively contaminated surfaces. Therefore, the decontamination performance of radioactive specimen taken from the commercial NPPs using HyBRID coupled with SP process was investigated.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2017

Large Eddy Simulation of turbulent flow in rod bundle with spacer grids

Pressurized water nuclear reactor fuel assemblies are made of fuel rods held by spacer grids, which are the core components of the nuclear power plants. Hot fuel rods are cooled by the coolant water flow, and the spacer grids with mixing vanes distort the mainstream, producing the swirling flow which can enhance the heat transfer to improve CHF (Critical Heat Flux). Heat transfer between fuel rods and coolant is a crucial part of thermal-hydraulic analysis. For the purpose to improve the thermal hydraulic performance of the fuel assemblies, many experiments have been conducted to understand the flow structures and three-dimensional behaviors of the turbulence in the rod bundle sub-channels and then to optimize the design of the spacer grid. In the experimental work, which requires lots of efforts to get reliable data, LDA (Laser Doppler Anemometry) and PIV (Particle Image Velocimetry) are generally used to measure the velocity profile and turbulent intensity in the sub-channels. Computational Fluid Dynamics (CFD) is another useful way to study the characteristics of the detailed flow condition and heat transfer phenomena in the subchannels. The main issue of the single-phase CFD is to accurately predict the flow and turbulence characteristics. RANS method is generally used to simulate the induced lateral velocity and the heat transfer process. With the development of computer technology, some more precise but previously time-consuming turbulent models, such as LES (Large Eddy Simulation), are gradually used for the simulation of rod bundle flow with current advanced CPU resources. Ikeno and Kajishima explained the decay of swirling turbulent flow in the rod bundle using LES method. Benhamadouche acquired the turbulence induced forces on the central rod of the rod bundle through performing the LES of the flow downstream the spacer grid. Chang and Tavoularis conducted hybrid simulations with SAS SST and LES of a 5 x 5 rod bundle, and several vortices were identified in the vicinity of the vanes and the generation mechanisms, evolution progress and influence on the velocities of these vortices were clarified. In the paper we present CFD simulations of turbulent flow in rod bundle sub-channels with spacer grids of two different experiments. One is MATiS-H benchmark experiment with scaled spacer grids, which is conducted by KAERI, and is used as a CFD benchmark test, and the other is a 5 x 5 rod bundle with typical spacer grids using PIV technology, with the cooperation of Harbin Engineering University. The main objectives of the calculations are to validate LES methodology by the benchmark experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Comparison of Error Rate Depending on Operator Expertise and Simulator Complexity

This paper analyzes operator's error rate from experiments, depending on the expertise and simulator complexity. This study uses the Rancor Microworld, a simplified simulator developed by INL, and the Compact Nuclear Simulator (CNS), a less simplified simulator developed by the Korea Atomic Energy Research Institute (KAERI). The error rates were measured from simulation data of a total of 72 participants, and the collected error rate data were analyzed using analysis of variance (ANOVA) test.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Summary Report of the FY24 DOE Contributions to the GIF VHTR CMVB

The Generation-IV Forum (GIF) Very-High-Temperature Reactor-Computational Methods Validation and Benchmark (VHTR-CMVB) initiative, involving organizations from Korea Atomic Energy Research Institute (KAERI) (South Korea), Institute of Nuclear and New Energy Technology of Tsinghua University (INET) (China), U.S. Department of Energy (DOE) (U.S.), Joint Research Centre (JRC) (Europe), and Japan Atomic Energy Agency (JAEA) (Japan), is dedicated to the verification and validation of tools for High-Temperature Gas-Cooled Reactors (HTGRs) analysis, using data shared by Computational Methods Validation and Benchmark (CMVB) signatories. For FY24, the US DOE CMVB has committed to several critical activities. Under WP1, led by the US, the integration of the High Temperature Gas Cooled Reactor - Pebble-Bed Module (HTR-PM) Phenomena Identification and Ranking Table (PIRT) into the comparative PIRT is progressing, with a new draft of the comparison tables issued earlier this year and currently being utilized by INET for their contribution. Neutronic validation efforts under WP3 include the preparation of the burnup analysis benchmark, preliminary calculations, and the development of reference models and results. In WP2, a validation exercise for hot gas mixing in the lower plenum of HTR-PM is in progress, using experimental data from INET (China) to validate modeling approaches. A model of the experimental facility has been developed using StarCCM+, with initial calculations slated for presentation at the GIF CMVB meeting this fall. Another WP2 activity focuses on validating numerical models for air-cooled Reactor Cavity Cooling System (RCCS) with experimental data from the Wisconsin Madison RCCS facility. A high-fidelity model, developed using NEK-RS, is currently being validated with available data from a low power forced convection test. These efforts are aimed at enhancing and confirming the accuracy of HTGR analysis tools, ensuring their alignment with experimental data and regulatory requirements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Summary Report Of The FY25 Computational Fluid Dynamics Verification And Validation Exercises In The Advanced Reactor Technologies - Gas-cooled Reactor Program

Verification and Validation (V&V) of numerical tools is critical for ensuring reasonable predictions during design, safety analysis, and licensing. Recent work in the Advanced Reactor Technologies - Gas-cooled Reactor (ART-GCR) program has focused on V&V of common Computational Fluid Dynamics (CFD) tools that are used within the Untied States. This report presents an update on these CFD V&V activities. These Generation IV Forum (GIF) Very High Temperature Reactor (VHTR) Computational Methods, Validation, and Benchmarking (CMVB) is an international organization dedicated to the verification and validation of High Temperature Gas-Cooled Reactor (HTGR) simulation tools. Participation in the CMVB provides additional value to the V&V activities, as it allows for access to a wider range of data, and provides valuable benchmarking exercises. Three HTGR phenomena are targeted: Reactor Cavity Cooling System (RCCS) performance, core bypass flow, and lower plenum mixing. Simulations of the University of Wisconsin-Madison (UW-Madison) RCCS facilities are performed with Reynolds Averaged Navier-Stokes (RANS) in StarCCM+. Results are compared for both forced and natural convection conditions, with both exhibiting good agreement with experimental measurements. The Idaho National Laboratory (INL) matched index of refraction (MIR) and Korean Atomic Energy Research Institute (KAERI) bypass flow expeirments are used to validation CFD predictions of bypass flow. Simulations are performed with RANS in StarCCM+ and with Large Eddy Simulation (LES) in NekRS. Finally, preliminary simulations of the Institute of Nuclear and New Energy Technology (INET) lower plenum mixing facilities are presented. Initial work has developed models with LES, RANS, and porous media models. These preliminary models are presented and compared to each other to gauge differences in predictions with each of the three methods.

and Benchmarking (CMVB)↗
DOE OSTI2025

Validation of RANS Models for Predicting Bypass Flow in Prismatic VHTRs

In prismatic Very-High Temperature Reactors (VHTRs), gaps occur between neighboring fuel assemblies referred to as bypass gaps. This work aims to measure the flow distribution between the bypass gaps between neighboring assemblies and the coolant channels within the assemblies. Computational fluid dynamics (CFD) was done using Reynolds-Averaged Navier Stokes (RANS) modeling in STAR-CCM+ to assess the ability of RANS models to accurately capture this phenomena. The RANS results were compared to experimental worked performed by Seoul National University (SNU) and Korea Atomic Energy Research Institute (KAERI) for validation. The CFD results showed good comparison with the experimental data available from SNU and flow split was characterized for multiple configurations, inlet mass flow rates, and bypass gap sizes.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗