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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Plenum plate standoff spools for separating a pair of plenum plates in a nuclear reactor

A nuclear reactor has first and second plenum plates disposed in a pressure vessel. Both plenum plates have a plurality of apertures. The second plenum plate is parallel to the first plenum plate. A fuel element includes a fuel element coolant flow tube which extends through aligned apertures of the parallel plenum plates. A fuel element standoff spool is disposed about a portion of the fuel element coolant flow tube which is located between the plenum plates. The nuclear reactor is also usable in nuclear thermal propulsion.

Inman, James B.↗
DOE OSTI2018

Experimental and Computational Investigations of Plenum-to-Plenum Heat Transfer and Gas Dynamics under Natural Circulation in a Prismatic Very High Temperature Reactor

All the goals and the objectives set for the project were successfully executed and achieved and all the milestones have been successfully completed. The results that have been obtained for the first time advance the scientific and engineering knowledge and understanding of the plenum-to plenum natural convection of prismatic block nuclear reactors that is encountered during accident or abnormal operation. These have been accomplished by developing and implementing for the first time unique and flexible scaled-down separate and integrated effects experimental plenumto- plenum facility (P2PF) with dual channels at this time that has been equipped with sophisticated measurement techniques integrated in a novel way on the heated and cooled channels. The unique facility is an asset now that can be extended to research multiple channels and to study the effects of hot plumes in the plena for future projects if funding will be available. It can also be modified to research natural convection of pebble bed reactors. Hence, it complement the HTTF at Oregon State University. However, in this study, heat transfer coefficients from the inner wall surface to the flowing gas (both helium and air were used) and the radial temperature and gas velocity profiles have been measured and investigated along the height of the heated and cooled channels using in house developed wall flush mounted heat transfer probes, thermocouple with in house developed adjuster for radial movement with 1 mm increment inside the channel and hot wire anemometry with also in house developed adjuster for 1 mm radial movement inside the channel, respectively. Also advanced tracer technique has been developed to quantify also for the first time the dispersion of the gas dynamics of the hot and cold channels. The research has provided new knowledge and new benchmarking data that can be used to validate computational fluid dynamics (CFD) codes with conjugate heat transfer. The work and its results that have been performed within the budget have demonstrated their superior technical effectiveness and high economic feasibility to perform needed studies for safety analysis and assessment at least cost for these types of gas cooled very high temperature 4th generation nuclear reactors. Accordingly, the results obtained in this project and the unique facility and techniques that have been developed will benefit greatly the public by advancing the technology of the prismatic block nuclear reactors toward commercialization and to ensure they will be designed and operated safely by utilizing the obtained knowledge and having well validated CFD simulations integrated with heat transfer computations

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Buoyancy-driven air flow within plenum-to-plenum facility down-comer channel

Reliable measurements of thermal and velocity fields were experimentally investigated under different natural circulation intensities in a dual channel facility designed and developed with a representative geometry of prismatic modular reactor core. Experiments were conducted under steady state conditions to obtain statistically stationary time series signals of temperature and velocity. Insignificant radial temperature variation was observed along the down-comer channel implying that the majority of supplied heat is removed through the upper plenum. Velocity results obtained emphasize the delicate nature of natural circulation phenomena in terms of flow destabilization, and recirculation penetration length. Observed air velocity distribution reveals that downward velocity is proportional to the extent of cooling applied to the setup. The calculated values of the dimensionless group Froude number ( Fr ) in the range of 2 to 10 indicate that air is dominated by momentum and is exiting the channel as a jet. Quantification of turbulent intensities implies that flow destabilization is noticeable along the down-comer channel particularly for the case of high cooling intensity. For 5 °C cooling water temperature, turbulent intensity ($v$ rms / $v$) peaks at the mid-channel (z/L = 0.5) reaching a value of 0.5. This value of turbulent intensity is much higher than the inlet (z/L = 0.96) and outlet (z/L = 0.04) turbulent intensities of 0.2 and 0.06 respectively. This destabilization is because of the flow reversal and heat conduction effects through stainless steel flanges connecting both channels. In conclusion, current results provide detailed velocity and temperature variation which can be useful for validating computational fluid dynamics codes.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2023

Investigation of thermal hydraulic behavior of the High Temperature Test Facility's lower plenum via large eddy simulation

A high-fidelity computational fluid dynamics (CFD) analysis was performed using the Large Eddy Simulation (LES) model for the lower plenum of the High–Temperature Test Facility (HTTF), a ¼ scale test facility of the modular high temperature gas-cooled reactor (MHTGR) managed by Oregon State University. In most next–generation nuclear reactors, thermal stress due to thermal striping is one of the risks to be curiously considered. This is also true for HTGRs, especially since the exhaust helium gas temperature is high. In order to evaluate these risks and performance, organizations in the United States led by the OECD NEA are conducting a thermal hydraulic code benchmark for HTGR, and the test facility used for this benchmark is HTTF. HTTF can perform experiments in both normal and accident situations and provide high-quality experimental data. However, it is difficult to provide sufficient data for benchmarking through experiments, and there is a problem with the reliability of CFD analysis results based on Reynolds–averaged Navier–Stokes to analyze thermal hydraulic behavior without verification. To solve this problem, high-fidelity 3-D CFD analysis was performed using the LES model for HTTF. It was also verified that the LES model can properly simulate this jet mixing phenomenon via a unit cell test that provides experimental information. As a result of CFD analysis, the lower the dependency of the sub-grid scale model, the closer to the actual analysis result. In the case of unit cell test CFD analysis and HTTF CFD analysis, the volume-averaged sub-grid scale model dependency was calculated to be 13.0% and 9.16%, respectively. As a result of HTTF analysis, quantitative data of the fluid inside the HTTF lower plenum was provided in this paper. As a result of qualitative analysis, the temperature was highest at the center of the lower plenum, while the temperature fluctuation was highest near the edge of the lower plenum wall. The power spectral density of temperature was analyzed via fast Fourier transform (FFT) for specific points on the center and side of the lower plenum. FFT results did not reveal specific frequency-dominant temperature fluctuations in the center part. It was confirmed that the temperature power spectral density (PSD) at the top increased from the center to the wake. The vortex was visualized using the well-known scalar Q-criterion, and as a result, the closer to the outlet duct, the greater the influence of the mainstream, so that the inflow jet vortex was dissipated and mixed at the top of the lower plenum. Additionally, FFT analysis was performed on the support structure near the corner of the lower plenum with large temperature fluctuations, and as a result, it was confirmed that the temperature fluctuation of the flow did not have a significant effect near the corner wall. In addition, the vortices generated from the lower plenum to the outlet duct were identified in this paper. It is considered that the quantitative and qualitative results presented in this paper will serve as reference data for the benchmark.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Design of an experimental facility with a unit cell test section for studies of the lower plenum in prismatic high temperature gas reactors

A proposed design for a Generation IV reactor is the very-high temperature reactor (VHTR), which employs helium as the primary coolant. Although many advantages exist with this design, a number of challenges remain. One such area of concern is the turbulent mixing of non-isothermal flows in the lower plenum, where an array of coolant jets collectively mix and traverse around a series of structural support posts before exiting the core. Neighboring jets in the lower plenum could experience temperature differences as high as 300–400 K, causing potential hot streaking in the support posts. Additional problems could occur further downstream if the level of non-uniformity in the temperatures is too high (e.g, inducing thermal stresses on gas turbine blades). Therefore, it is important that adequate tools exist to predict the turbulent mixing within the VHTR lower plenum. This task remains difficult to achieve due to the complex geometry and flow physics that dictate the mixing. Various modeling approaches can only be assessed using data acquired from a carefully designed experimental facility. This paper presents details of such a facility, where high-fidelity measurements can be acquired for a scaled portion of the lower plenum defined herein as a ‘‘unit cell”, which consists of a hexagonal array of jets whose flow is directed into a test section with support posts positioned between neighboring jets, and subjected to a crossflow. Through adjusting velocity and temperature conditions at the jet inlets, different locations within the lower plenum can be experimentally simulated. System response quantities (SRQs) include both temperature and velocity measurements, captured via multiple thermocouples embedded in each post, and particle image velocimetry (PIV), respectively. These SRQs provide assessment metrics for both the momentum and energy models present in current computational solvers, and therefore the experimental facility is a valuable source for validation data. Another benefit is a more comprehensive understanding of the thermal-hydraulic loading conditions and flow physics present in the VHTR lower plenum.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Counterflow heat exchanger with core and plenums at both ends

Here, this paper illustrates the morphing of flow architecture toward greater performance in a counterflow heat exchanger. The architecture consists of two plenums with a core of counterflow channels between them. Each stream enters one plenum and then flows in a channel that travels the core and crosses the second plenum. The volume of the heat exchanger is fixed while the volume fraction occupied by each plenum is variable. Performance is driven by two objectives, simultaneously: low flow resistance and low thermal resistance. The analytical and numerical results show that the overall flow resistance is the lowest when the core is absent, and each plenum occupies half of the available volume and is oriented in counterflow with the other plenum. In this configuration, the thermal resistance also reaches its lowest value. These conclusions hold for fully developed laminar flow and turbulent flow through the core. The curve for effectiveness vs number of heat transfer units (N tu ) is steeper (when N tu < 1) than the classical curves for counterflow and crossflow.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2019

Calculation of Neutron Damage to Sodium Coolant Inlet Plenum

The coolant inlet plenum is a life limiting component of a Sodium-cooled Fast Reactor (SFR). The inlet plenum is a fixed reactor component which receives sodium from the cold side of the sodium pool redan and directs it into the inlet nozzles of the fuel assemblies. Neutron leakage from the reactor core causes irradiation damage to the steel components of the inlet plenum leading to a reduction in strength. The life of the reactor is essentially tied to the acceptable strength degradation of the inlet plenum. Therefore, typical SFR designs employ a neutron reflector/shield as an integral feature of the fuel assembly. The role of the axial reflector/shield is to: 1) direct fast neutrons back to the core, and 2) attenuate the neutron flux to minimize neutron interaction with the inlet plenum. The longer the reflector and the greater the share of steel within the reflector design, the greater the attenuation. The axial reflector is heated by gamma rays from the core, and must have sufficient coolant penetrations to ensure adequate cool-ability. Also, the sodium coolant must pass through the reflector before reaching the fuel rods above it. The present work investigates varying reflector geometries in support of the Versatile Test Reactor (VTR), a SFR research reactor planned to restart fast reactor irradiation capabilities in the U.S. DOE laboratory complex. This work will demonstrate that the neutron damage to the inlet plenum is relatively insensitive to the number and arrangements of coolant channel penetrations through the reflector.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Effect of cold cap coverage and emissivity on the plenum temperature in a pilot-scale waste vitrification melter

Integrated computational fluid dynamics (CFD) models are being developed to model the complex physics occurring within the high-level waste melter for vitrification of legacy tank waste at the Hanford site. Here, we present a validation of the integrated CFD model by using data from two experimental runs in a pilot-scale melter. While the model uses several simplifying assumptions (such as constant heat sinks from a cooling jacket and inleakage of ambient air, steady state feed-to-batch conversion heat, and a cold cap model with a simplified shape), it closely predicts the molten glass (1150°C and 1175°C) and plenum temperatures (550°C) obtained from thermocouples during two pilot-scale tests, with an average cold cap coverage of 80%. Additional simulations were performed to explore the sensitivity of the predicted plenum temperatures to variations in cold cap coverage (fraction of melt surface covered by the glass batch) and batch emissivity. The plenum temperature was found to be in the range of 606°C when cold cap coverage decreased from 95% to 70%. Cold cap emissivity had a smaller effect, increasing the plenum temperature by as much as 179°C when cold cap emissivity increased from 0.2 to 0.8. Maintaining a high cold cap coverage without overfeeding is important for a sustained melter operation with high glass throughput. This work provides a tool for achieving that goal in terms of correlating the plenum temperature with the cold cap coverage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Performance Characteristics of a Rotating Detonation Combustor Exiting Into a Pressurized Plenum to Simulate Gas Turbine Inlet

Abstract The present study aims to experimentally characterize the performance of a rotating detonation combustion (RDC) system integrated with a pressurized downstream plenum to simulate the high-pressure inlet conditions of power-generating gas turbines. A thorough understanding of the operational behavior including wave mode behavior, static pressure profile along the combustor length, and dynamic features of pressure fluctuations is crucial for successful integration of RDC with the turbine. In this study, two RDC configurations are investigated, RDC with a constant area annulus and RDC with a converging nozzle. In both cases, the RDC flow exited into a plenum chamber kept at pressures varying from 155 kPa to 330 kPa. RDC was operated on methane and oxygen-enriched air to represent reactants used in land-based power generation. Experiments were conducted for the two RDCs configurations operated at three reactant mass flow rates (0.23, 0.32, and 0.46 kg/s). The RDC performance is characterized by time-averaged static pressure measurements, and wave velocity determined by ionization probes. In addition, dynamic pressure measurements were recorded both inside and near the exit of RDC channel to investigate wave interactions between RDC and downstream plenum. Results show that the RDC with the converging nozzle achieved superior performance while minimizing detrimental interactions with the reflected shock and/or acoustic waves from the downstream plenum.

Engineering↗
DOE OSTI2020

3D CFD Model Validation Using Benchmark Data of 1/16th Scaled VHTR Upper Plenum and Development of Wall Heat-Transfer Correlation For Laminar Flow

With support from the U.S. Department of Energy-Office of Nuclear Energy’s (DOE-NE’s) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, an effort has been pursued to support high-temperature gas-cooled reactor (HTGR) technology development and its modeling and simulation needs. There is a particular need for advanced modeling and simulation tools to predict thermal-fluid behavior in the nuclear reactor primary system, especially in the core and the lower and upper plena, during safety-related transients. In this report, two main such activities are presented relevant to the HTGRs: (1) three-dimensional (3D) computational fluid dynamics (CFD) validation using benchmark data from the upper plenum of Texas A&M University’s 1/16th scaled very-high-temperature gas-cooled reactor (VHTR), and (2) development of wall heat-transfer correlation for laminar flow in a wall-heated pipe. The CFD tool validation exercises can be helpful to choose the models and CFD tools to simulate and design specific components of the HTRGs such as upper plenum where jet mixing is a complex phenomenon. In a loss of forced circulation event, the laminar flow can be observed during the development of natural circulation flow. This work includes the development and validation of heat transfer correlations for laminar flow using the Nek5000 CFD code due to limited available experimental data for laminar flow conditions to guide low-order models (1D). In this report, the flow characteristics of a single isothermal jet discharging into the upper plenum was investigated using the Nek5000 Large-Eddy Simulation (LES) CFD tool. Several numerical simulations were performed for various jet-discharged Reynolds numbers ranging from 3,413 to 12,819. A grid-independent study was performed. The numerical results of mean velocity, root-mean-square fluctuating velocity, and Reynolds stress were compared against the benchmark data. Good agreement was obtained between simulated and measured data for axial mean velocities, except near the upper plenum hemisphere. The maximum predicted errors for axial mean velocities at various normalized coolant channel diameter heights of 1, 5, and 10 are 1.56%, 1.88%, and 3.82%, respectively. In addition, the predicted root-mean-square fluctuating velocity and Reynolds stress are qualitatively in agreement with the experimental data. The Nek5000 code was used to develop wall-heat transfer correlation for laminar flow in a cylindrical tube. Several simulations were performed for various Reynolds flow and wall-heat fluxes. A new heat transfer correlation was developed using data from Nek5000 simulation results and regression functions in Matlab. The developed heat transfer correlation is valid for various Reynolds flows from 200 to 2000. The predicted R² value for model fit was 0.875, which ensures that 87.5% of the model data lies on the Nek5000 data. Moreover, a machine learning (ML) tool was used to train and test the Nek5000 data. A good fit of the ML-based model was observed with the test data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Turbine airfoil with leading edge cooling passage(s) coupled via plenum to film cooling holes, and related method

A turbine airfoil includes a body including a wall defining pressure and suction sides, and a leading edge extending between the pressure and suction sides. A cooling circuit inside the wall of the body includes at least one of: a) a suction side to pressure side cooling sub-circuit including a first cooling passage(s) extending from the suction side to the pressure side around the leading edge to a first plenum, and a plurality of first film cooling holes communicating with the first plenum and extending through the wall on the pressure side; and b) a pressure side to suction side cooling sub-circuit including second cooling passage(s) extending from the pressure side to the suction side around the leading edge to a second plenum, and a plurality of second film cooling holes communicating with the second plenum and extending through the wall on the suction side.

Lacy, Benjamin Paul↗
DOE OSTI2019

Computational study of full-scale VHTR lower plenum for turbulent mixing assessment

Next-generation nuclear reactors are poised to efficiently provide reliable power at enhanced safety levels for many years to come. Among the fundamental designs for these reactors is the very-hightemperature reactor (VHTR), which employs helium as the primary coolant and has the potential of reaching elevated temperatures such that chemical processing (e.g., electrolysis for hydrogen production) is viable alongside traditional electricity generation. This and other next-generation reactors also represent a strategic bridge for ultimately transitioning from fossil fuel-dominated energy portfolios to one where renewable options abound. For the VHTR, the ultimate goal of constructing a new plant will be preceded by additional fundamental research aimed at generating trusted models for behavior prediction under normal and accident scenarios. This paper presents a detailed computational fluid dynamics simulation of the lower plenum, where hot coolant from the core mixes together in a turbulent fashion before traveling to power conversion equipment. Because the flow is expected to enter the lower plenum across a wide range of temperatures and velocities, concerns exist when the mixing is incomplete. The potential hot spots on cylindrical support pedestals and the uniformity of the main outlet are the primary metrics of interest in this study, along with flow velocities and temperatures at different locations within the lower plenum. Three locations are probed in detail and suggest a large range exists for approach velocities and temperatures. A large degree of stratification is also seen on the surfaces of the support pedestals, suggesting a facility should accommodate testing for such behavior. The characterizations presented provide the valuable data needed in order to design appropriate experimental testbeds, where scaled modeling can be carried out in a manner meaningful in predicting the full-scale behavior.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

CFD Analysis of Upper Plenum Flow for a Sodium-Cooled Small Modular Reactor

Upper plenum flow behavior is important for many operational and safety issues in sodium fast reactors. The Prototype Gen-IV Sodium Fast Reactor (PGSFR), a pool-type, 150 MWe output power design, was used as a reference case for a detailed characterization of upper plenum flow for normal operating conditions. Computational Fluid Dynamics (CFD) simulation was utilized with detailed geometric modeling of major structures. Core outlet conditions based on prior system-level calculations were mapped to approximate the outlet temperatures and flow rates for each core assembly. Core outlet flow was found to largely bypass the Upper Internal Structures (UIS). Flow curves over the shield and circulates within the pool before exiting the plenum. Cross-flows and temperatures were evaluated near the core outlet, leading to a proposed height for the core outlet thermocouples to ensure accurate assembly-specific temperature readings. A passive scalar was used to evaluate fluid residence time from core outlet to IHX inlet, which can be used to assess the applicability of various methods for monitoring fuel failure. Additionally, the gas entrainment likelihood was assessed based on the CFD simulation results. Based on the evaluation of velocity gradients and turbulent kinetic energies and the available gas entrainment criteria in the literature, it was concluded that significant gas entrainment is unlikely for the current PGSFR design.

computational fluid dynamics↗
DOE OSTI2016

Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensity peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by comparing it with benchmark data. The experimental data gave the same pattern as the benchmark data. A turbulent single buoyant jet study was performed for the case of LOFC in the upper plenum of scaled VHTR. Time-averaged profiles show that 3,000 frames of images were sufficient for the study up to second-order statistics. Self-similarity is an important feature of jets since the behavior of jets is independent of Reynolds number and a sole function of geometry. Self-similarity profiles were well observed in the axial velocity and velocity magnitude profile regardless of z/D where the radial velocity did not show any similarity pattern. The normal components of Reynolds stresses have self-similarity within the expected range. The study shows that large vortices were observed close to the dome wall, indicating that the geometry of the VHTR has a significant impact on its safety and performance. Near the dome surface, large vortices were shown to inhibit the flows, resulting in reduced axial jet velocity. The vortices that develop subsequently reduce the Reynolds stresses that develop and the impact on the integrity of the VHTR upper plenum surface. Multiple jets study, including two, three and five jets, were investigated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

A New Fast Fluid Dynamics Model for Data-Center Floor Plenums

A raised-floor plenum is often used to distribute cooling airflow in data centers. Traditional Computational Fluid Dynamics (CFD) can accurately model airflow distribution in plenums, but it is expensive (computationally and financially) and is, therefore, often not accessible to the data center designer or operator. This paper presents an alternative CFD methodology based on Fast Fluid Dynamics (FFD), which is faster, more parallelizable, and simpler to program. The improvement in computational speed afforded by FFD is particularly important for multiple-iteration optimization, transient simulations, or applications for which near-real-time performance is required. Here, the potential speed improvements of FFD have been widely documented; however, it has been generally accepted that this benefit is accompanied by a modest loss of accuracy relative to traditional CFD. We show here that this is not necessarily true. Plenum airflow predictions by our FFD implementation, which utilizes a first-order upwind rather than Semi-Lagrangian advection scheme of previous research, are extremely close to those predicted by traditional CFD - provided the same turbulence model and computational grid are used consistently.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2025

CFD Simulations of Lower Plenum Mixing

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

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Use of scaled BWR lower plenum boron mixing tests to qualify the boron transport model used in TRACG

In 2001 GEH applied best estimate methods combined with a statistical methodology to determine upper bound limits for key licensing parameters for anticipated operation occurrence (AOO) transient and anticipated transients without scram (ATWS) overpressure analyses for operating Boiling Water Reactors (BWRs). The methodology was subsequently extended for ESBWR AOO, ATWS, loss of coolant, and stability analyses. GEH is extending the methodology to long-term ATWS analyses for the operating BWRs. A long-term ATWS scenario uses injection of borated water to achieve reactor shutdown. Predicting the mixing and transport of boron is important for calculating the impact on the key licensing parameters. For the many operating BWRs where the denser boron solution is injected into the lower plenum, stratification may occur, delaying boron transport to the core region. CFD modeling can be used to model the stratification and mixing of the boron solution, but such calculations are extremely computer intensive and not cost effective; therefore, a more-empirical approach supported by a theoretical scaling of the dominant phenomena and backed by test data and benchmark calculations is used. The paper presents the TRACG lower plenum boron transport model qualification effort. The scaling basis used to implement the TRACG boron transport model for BWR applications is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗