High-temperature Test Facility Research Articles

RELAP5-3D validation studies based on the High Temperature Test facility

In the spring and summer of 2019, experiments were conducted at the High Temperature Test Facility (HTTF) that form the basis of an upcoming high-temperature gas-cooled reactor (HTGR) thermal hydraulics (T/H) benchmark. HTTF is an integral effects test facility for HTGR T/H modeling validation. This paper presents RELAP5-3D models of two of those experiments: PG-27, a pressurized conduction cooldown (PCC); and PG-29, a depressurized conduction cooldown (DCC). These models used the RELAP5-3D model of HTTF originally developed by Paul Bayless as a starting point. The sensitivity analysis and uncertainty quantification code, RAVEN was used to perform calibration studies for the steady-state portion of PG-27. We developed four PG-27 calibrations based on steady-state conditions. These calibrations all used an effective thermal conductivity equal to 36 % of the measured thermal conductivity, but they differed with respect to the frictional pressure drops and radial conduction models. These models all captured the trends in steady-state temperature distributions and transient temperature behavior well. All four calibrations show room for improvement in predicting the transient temperature rise. The smallest error in temperature rise during the transient was a 21 % underprediction, and the largest was a 48 % underprediction. The errors in transient temperature rise are largely a result of a mismatch in power density between the RELAP5-3D model and the experiment due to the location of active heater rods along the boundary between heat structures in the model. The best of these calibrations was applied to PG-29 to model the DCC. Once again, temperatures during the transient were underpredicted but trends in temperature were captured. The RELAP5-3D model captured trends in the data but could not reproduce measured temperatures exactly. This result is not attributed to deficiencies in the experimental data or to RELAP5–3D itself. Rather, this result likely arises due to the some of the assumptions and decisions made when the RELAP5-3D model was first developed, prior to the execution of HTTF experiments. An agreement in prediction of temperature trends but challenges reproducing HTTF temperatures within measurement uncertainty is consistent with previous analyses of HTTF in the literature. Future RELAP5-3D validation activities centered around HTTF may be able to provide greater insight into the code’s capabilities for HTGR modeling with a more finely nodalized model.

ASYST4.1 validation for gas cooled SMR applications using the HTTF experiment

Small Modular Reactors (SMRs) based on High Temperature Gas Cooled Reactors (HTGR) concepts are being considered for remote communities in Canada (power ∼ 5 MWe or less) and elsewhere. One such design is Micro-Modular Reactor (MMR) from USNC (Ultra-Safe Nuclear Corporation). It uses helium as primary coolant and molten salt as intermediate coolant and for thermal energy storage (TES). The molten salt can then be used to generate steam on-demand to match the local communities’ energy requirements. Reliable computational tools are needed for thermal hydraulic analysis of the integrated nuclear and molten salt systems.ASYST4.1 (Advanced SYStem Thermal system) is one such tool which has been recently validated for Solar Salt (40 % NaNO3 + 60 %KNO3) applications and has capability to simulate the helium coolant and reactor core behaviour. To enable the code to model the primary and intermediate loops of MMR there is a need to validate the code for helium coolants and its capabilities to simulate lateral conduction, natural circulation, and radiative heat transfer. The HTTF (High Temperature Test Facility) experiment PG-26, a Depressurized Conduction Cooldown transient, is used for this validation exercise. The results are compared with those of the experiment and those from RELAP5-3D calculations. The reference case and sensitivity analyses for the important parameters are presented.Mass flow rate and temperatures in the primary loop are compared for the steady-state and for the transient PG-26. For the initial assessment ASYST4.1 predictions are found in good agreement with RELAP5-3D, but these results deviate from the experiment. Significant improvement in simulation results is observed when simulations consider the entire trajectory of the warm-up period and of the experiment, indicating that the experiment was not in a steady-state condition at the time of the depressurization test. This is an important factor to be considered in all HTTF cases going forward.

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.

Impact of Temperature Ratio on Overall Cooling Performance: Low-Order-Model-Based Analysis of Experiment Design

Abstract This paper describes low-order-model-based analysis of the design of an experiment to be used for parametric studies of adiabatic film and overall cooling effectiveness for fully cooled systems (internal and film) under wide ranges of mainstream-to-coolant temperature ratio variation, in the range 0.50 < T0m/T0c < 2.30. The purpose is to improve understanding of—and validation of—the scaling process from typical rig conditions to engine conditions. We are primarily interested in the variation in overall effectiveness when the controlling non-dimensional groups change in a natural co-dependent way with changes in temperature ratio: that is, the practical situation of interest to engine designers. We distinguish this from the situation in which individual non-dimensional groups are varied in isolation: a situation that we believe is essentially impossible to meaningfully approximate in practice, despite a body of literature purporting to do the same. Design and commissioning data from a new high temperature (600 K) test facility is presented, with detailed uncertainty analysis. We show (using a low-order model) that a typical nozzle guide vane which at engine conditions (TR = 2.00) would have overall cooling effectiveness of 0.450, would be expected to have overall effectiveness of 0.418 at typical rig conditions (TR = 1.20). That is, typical scaling from engine-to-rig result is −7.1% and typical scaling from rig-to-engine is +7.7%. This result is important for first order estimation of overall cooling performance at engine conditions.

Verification and validation of a lower plenum mixing test at the OSU High Temperature Test Facility

Verification and validation of a lower plenum mixing test at the OSU High Temperature Test Facility

Code Benchmark of the HTTF Pressurized Conduction Cooldown Test Using SAM

The High Temperature Test Facility (HTTF) at Oregon State University is an integral system test facility to simulate postulated reactor transients of prismatic high-temperature gas-cooled reactors(HTGRs). A series of test campaigns was launched, providing abundant test data that could be used to benchmark reactor system analysis codes like the System Analysis Module (SAM). In this study, a SAM model of the facility is developed based on the two-dimensional (2D) ring model approach. All components including the ceramic matrix, graphite heaters, helium coolant channels, core barrel, upcomer, pressure vessel, and reactor cavity cooling system are modeled as concentric cylindrical rings. The model is used to simulate one of the benchmark problems—Pressurized Conduction Cooldown (PCC)—within the scope of the Organisation for Economic Co-operation and Development Nuclear Energy Agency International HTTF Benchmark. The simulations consist of two parts. In the first part, operating and boundary conditions as well as thermophysical properties of materials are specified for the benchmark problem. Results from the first part will be used in code-to-code comparison. In the second part, the SAM model is used to simulate Test PG-27, which is the first PCC test carried out in the HTTF, with only two of the ten heater banks activated. The results in the second part are used for code-to-data comparison. Because the helium coolant flow rate is not measured in this facility, it is estimated using the input power and inlet/outlet coolant temperatures. Additionally, radial heat flow in the ceramic blocks is complicated by hundreds of cylindrical coolant channels and heater rods embedded in them. As such, it is necessary to deduce an effective thermal conductivity for the ceramic to analyze the core thermal behavior. SAM predictions of the helium coolant and ceramic temperatures are compared with test data measured in three equivalent sectors. Overall, the SAM results agree reasonably well with test data within the variation of data among the three sectors, which demonstrates SAM’s capability in capturing transient effects in HTGR using the simplified 2D ring model.

High Temperature Test Facility sensitivity and calibration studies to inform OECD-NEA benchmark calculations

The High Temperature Test Facility (HTTF) is a thermal hydraulics test facility for block-type high-temperature gas-cooled reactors (HTGRs) built to provide data to validate HTGR modeling tools. Several experiments were conducted at HTTF in the spring of 2019 that serve as the basis for an upcoming Organization for Economic Cooperation and Development – Nuclear Energy Agency HTGR thermal hydraulics benchmark. We present sensitivity studies of steady state and both the pressurized and depressurized conduction cooldown (PCC and DCC) transients in HTTF in this paper. These studies identified that uncertainties in the thermal conductivity of the ceramic blocks making up the HTTF core are significant drivers of peak temperatures in steady state, PCC, and DCC. The emissivity of the heater rods and core blocks also provide significant control over heater rod temperatures. We also performed calibration studies based on HTTF Experiment PG-26, which was intended to serve as the basis for the DCC benchmark problem. We found that no calibration was able to reproduce PG-26 peak block temperatures within an average of 30%, but by taking the best-estimate PG-26 calibration and applying it to a similar experiment—PG-29—we were able to reproduce the peak block temperatures within an average of 6.2% error. Based on this nearly fivefold improvement in peak block temperature predictions from PG-26 to PG-29 and confounding factors in PG-26 that contributed to previous difficulties modeling the experiment in RELAP5-3D and SAM we conclude that PG-26 data are insufficient for developing a DCC benchmark and recommend using PG-29 as the benchmark experiment instead.

Developing PCC and DCC integral effects test experiments at the High Temperature Test Facility

Among the Next-Generation Nuclear Plant (NGNP) designs, the High-Temperature Gas-Cooled Reactors (HTGRs) are very attractive, due to their inherent safety features, high power conversion efficiency, and potential of providing high-temperature process heat. To perform a thorough safety study and to license these types of reactors, sufficient information needs to be provided about the phenomena that occur during accident scenarios. While several experimental research efforts have been dedicated in the past to investigate accident scenarios, knowledge gaps still exist in the phenomena characteristic of pressurized and depressurized conduction cooldown (PCC/DCC) transients as well as for normal operation scenarios. This paper summarizes the Oregon State University High Temperature Test Facility (HTTF) test matrix, experimental campaign, and selected tests results. High Temperature Test Facility is a scaled Integral Test Facility (IET) that is capable of mimicking scaled dimensions and operational conditions of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The goal of the High Temperature Test Facility is to provide experimental data on the DCC, PCC and normal operating scenarios of the reference Modular High-Temperature Gas Cooled Reactor design. The DCC, PCC, mixing, heat up and cooldown tests described in this paper were performed at prototypical Modular High-Temperature Gas Cooled Reactor temperatures, scaled initial pressure conditions (∼200 kPa), and thermal power input of less than 70 kW. Presented test data show temperature distributions in the High Temperature Test Facility core, upper plenum, cross duct, or lower plenum. Based on these temperature profiles attempts to investigate stratified flow, natural convection flow, heat up, cooldown and mixing phenomena are made. Furthermore, this paper evaluates the performed test campaign in the light of the Very High Temperature Gas-cooled Reactor Phenomena Identification and Ranking Table (PIRT) and proposes experiments to complement the existing PCC/DCC testing database for the validation of the thermal-hydraulic codes.

PERFORMANCE INVESTIGATIONS OF THE ENERGY DISCHARGE CHARACTERISTICS OF HIGH-TEMPERATURE PHASE CHANGE MATERIAL CAPSULES

Thermal energy storage using phase change material (PCM) capsules is a prevalent method used in high-temperature applications such as steam generation and concentrated solar power. However, the research works on the development of latent heat storage PCM encapsulations are limited. In the present study, experimental and numerical investigations conducted on high-temperature cylindrical encapsulations filled with sodium nitrate that were using air as the heat transfer fluid are presented. A modification of the cylindrical capsule known as the annular capsule is proposed to improve the heat transfer characteristics inside the PCM by providing a heat transfer surface at the center of the capsule. Experimental investigations were conducted utilizing a high-temperature (723 K) test facility to determine the discharging characteristics of both capsule designs in the temperature range between 603 and 473 K. A two-dimensional axisymmetric numerical model is developed and validated with the results obtained from the experiments. The performances of both the basic and annular capsules are compared, and it is observed that the discharging time reduced by 24% for the annular capsule. The heat transfer enhancement occurs due to an increase in the contact surface area and the improvement in flow characteristics of the air.

Simulation of the High Temperature Test Facility (HTTF) Core Using the 2D Ring Model with SAM

A two–dimensional ring model is developed with SAM to model the core of the High Temperature Test Facility (HTTF) at the system level. The ring model simplifies the complex structure of the HTTF core by converting the hexagonal rows of heaters and flow channels into layers of concentric annular rings. The ring model is first compared against a three–dimensional (3D)–one–dimensional (1D) model where the solid structures are fully resolved in three dimensions while the fluid structures are modeled as 1D flows. Comparison between the 3D–1D and the ring models shows that the latter can predict major parameters reasonably well under steady–state normal operating conditions, but the heater temperatures are under predicted. Adjustment is made to the effective thermal conductivity of the ceramic core of the ring model to improve the heater temperature predictions. The ring model is also used to simulate a transient pressurized conduction cooldown condition and is benchmarked with the experimental data from the HTTF Test PG–27. Good agreement is obtained between the experimental data and the predictions by the ring model.

Coupled simulations for prismatic gas-cooled reactor

The development and deployment of gas-cooled reactors (GCRs) require flexible modelling and simulation tools for a detailed understanding of the system behavior in normal operation and accidental scenarios. Amongst the identified safety-relevant scenarios for the gas-cooled reactors, outlet plenum flow distribution during normal operations was ranked to be of high importance with a low knowledge level in the phenomenon identification and ranking table (PIRT). The present paper contributes to the development of accurate simulation models using stand-alone and coupled codes for the GCR lower plenum region, which are needed to address nuclear reactor safety issues early enough so that delays in licensing and facility construction can be minimized. For the first time in this paper, an assessment of the predictive capabilities of coupled system-thermalhydraulics and CFD codes against heat transfer measurements in the lower plenum of a prismatic block GCR is reported. The lower plenum mixing experiments of the integral high temperature test facility at the Oregon State University (OSU HTTF) were simulated in the stand-alone and coupled mode to predict the lower plenum mixing. The coupled-code analysis used a 3D CFD code STAR-CCM + and a 1D thermalhydraulics code RELAP5-3D to account for system-level feedback on the flow and heat transfer for lower plenum mixing occurring during the normal operation at the reduced power and/or mass flow rate. For one of the simulated OSU test conditions, the existing models in STAR-CCM + and RELAP5-3D predicted the measured core temperatures reasonably well at two different elevations, suggesting that the available models are adequate for such an analysis. Through this study, a better understanding of the suitability of the models in the CFD and system thermalhydraulics (TH) codes is gained for more accurate prediction of flow and heat transfer in GCRs in anticipation of the eventual need to perform safety and licensing analyses.

Pressure drop and heat transfer characteristics of nitrate salt and supercritical CO2 in a diffusion-bonded heat exchanger

A lab-scale 316/L stainless steel diffusion-bonded heat exchanger was examined within a coupled high-temperature test facility consisting of an atmospheric-pressure nitrate salt loop and a supercritical carbon dioxide (sCO2) loop operating between 10 and 16 MPa. Pressure loss measurements collected for flow rates up to 0.55 kg/s of Solar Salt and up to 0.6 kg/s of carbon dioxide were used to generate friction factor correlations for the respective circular and semicircular zig-zag flow passages. Heat transfer measurements were also performed over a wide range of conditions to produce Nusselt correlations for the CO2 and nitrate salt geometries. The Nusselt number is deduced from a heat transfer resistance network and overall conductance measurements collected across the operating envelope of the coupled sCO2 – nitrate salt test facility. The CO2 side correlation showed excellent agreement with existing correlations in the literature. The salt side, having nearly circular channels with a different zig-zag angle and hydraulic diameter, yielded a new correlation applicable to the laminar flow regime. To examine the similitude of the friction correlations, pressure drop testing with water was performed on both sides of the component as well. Finally, results from non-destructive examination are discussed to provide insights regarding inspection methods and draining performance of high-temperature liquids in compact heat exchangers.

CFD modeling of the OSU High Temperature Test Facility inlet plenum flow distribution during normal operation

This paper investigates the fundamental physical phenomena associated with internal coolant flow distribution in the upper plenum of a prismatic Very High Temperature Reactor (VHTR) during normal operation. Previous studies have revealed the importance of complex flow distribution in the inlet plenum with the potential to generate low velocity or stagnation zones that can subsequently lead to the formation of hot spots in the reactor core and hot streaks in the lower plenum. It is therefore of interest to ensure that coolant is evenly distributed when entering multiple reactor coolant channels. Non-uniformity in the flow distribution is assessed for the reference Oregon State University High Temperature Test Facility (HTTF) case. The HTTF is a reduced scale model (1:4 in height and diameter) of the Modular High Temperature Gas-Cooled Reactor (MHTGR). The developed CFD model reflects complete and detailed geometry of the HTTF upper plenum along with upcomer and metallic core support structure (MCSS). CFD results include parametric studies performed for circulator mass flow rate variation and mesh sensitivity analysis, generated using Siemens Starccm+ software. Outcomes also include variable pressure and temperature boundary conditions impact on the flow distribution.

Velocimetry during depressurized conduction cooldown events in the HTTF

Abstract The High Temperature Test Facility (HTTF) is a quarter-scale integral-effect facility designed to study pressurized and depressurized conduction cooldowns (PCC and DCC, respectively) in high temperature, gas-cooled reactors (HTGR). This study focuses on DCC and aims to characterize the gas exchange between the reactor core and the reactor cavity to assess the risk for air ingress and subsequent potential for natural circulation core cooldowns. High-resolution velocity profile measurements are performed at the coolant pipe break location over a long time-scale using molecular tagging velocimetry. This non-intrusive, laser-based technique is applied for the first time in the field of nuclear thermal hydraulics, providing new insights for understanding the underlying phenomena at play during a DCC event. Such data are also valuable for validation of numerical simulations such as Reactor Excursion and Leak Analysis Program (RELAP). Results are reported here for various gas mixtures of helium (the coolant) and nitrogen (surrogate for air) to study the effect of density ratio. The short transient (30 s) lock-exchange phase is well captured with helium flow velocity between 0.8 and 2 m/s. The flow is shown to persist over much longer time-scales (hours) at a velocity of about 0.2 m/s, a likely consequence of gas mixing in the hot leg and in the lower plenum of the reactor. Flow visualization at the exit of the hot leg shows shear instabilities, which supports the hypothesis of significant mixing.

Development and Application of Molecular Tagging Velocimetry for Gas Flows in Thermal Hydraulics

Molecular tagging velocimetry (MTV) is a nonintrusive velocimetry technique based on laser spectroscopy. It is particularly effective in challenging gas flow conditions encountered in thermal hydraulics where particle-based methods such as particle image (or tracking) velocimetry do not perform well. The main principles for designing and operating this diagnostic are presented as well as a set of gases that have been identified as potential seeds. Two gases [H2O and nitrous oxide (N2O)] have been characterized extensively for thermodynamic conditions ranging from standard temperature and pressure to environments encountered in integral effects test (IET) facilities for high-temperature gas reactors. A flexible, modular, and transportable laser system has been designed and demonstrated with H2O and N2O seed gases. The laser system enables determining the optimum excitation wavelength, tracer concentration, and timing parameters. Velocity precision and thermodynamic domain of applicability are discussed for both tracers. The spectroscopic nature of the diagnostics enables one to perform first-principle uncertainty analysis, which makes it attractive for validating numerical models. Molecular tagging velocimetry is demonstrated for two flows. First, in blowdown tests with H2O seed, the unique laser system enables one of the largest dynamic ranges reported to date for velocimetry: 5000:1 (74 dB). N2O-MTV is then deployed in situ in an IET facility, i.e., the High-Temperature Test Facility at Oregon State University, during a depressurized conduction cooldown (DCC) event. Data enable researchers to gain insights into flow instabilities present during DCC. Thus, MTV shows a strong potential to gain a fundamental understanding of gas flows in nuclear thermal hydraulics and to provide validation data for numerical solvers.

Particle Deposition by Thermophoresis Under High-Temperature Conditions in a Helium Flow

The continuous generation of graphite dust particles in the core of a high-temperature reactor (HTR) is one of the key challenges of safety during its operation. The graphite dust particles emerge from relative movements between the fuel elements or from contact to the graphitic reflector structure and could be contaminated by diffused fission products from the fuel elements. They are distributed from the reactor core to the entire reactor coolant system. In case of a depressurization accident, a release of the contaminated dust into the confinement is possible. In addition, the contaminated graphite dust can decrease the life cycle of the coolant system due to chemical interactions. On one hand, the knowledge of the behavior of graphite dust particles under HTR conditions using helium as the flow medium is a key factor to develop an effective filter system for the discussed issue. On the other hand, it also provides a possibility to access the activity distribution in the reactor. The behavior can be subdivided into short-term effects like transport, deposition, remobilization and long-term effects like reactions with material surfaces. The Technische Universität Dresden has installed a new high-temperature test facility to study the short-term effects of deposition of graphite dust particles. The flow channel has a length of 5 m and a tube diameter of 0.05 m. With helium as the flow medium, the temperature can be up to 950 °C in the channel center and 120 °C on the sample surface, the Reynolds number can be varied from 150 up to 1000. The particles get dispersed into the accelerated and heated flow medium in the flow channel. Next, the aerosol is passing a 3 m long adiabatic section to ensure homogenous flow conditions. After passing the flow straightener, it enters the optically accessible measurement path made from quartz glass. In particular, this test facility offers the possibility to analyze the influence of the thermophoretic effect separately. For this, an optionally cooled sample can be placed in the measuring area. The thickness of the particle layer on the sample is estimated with a three-dimensional laser scanning microscope. The particle concentration above the sample is measured with an aerosol particle sizer (APS). Particle image velocimetry (PIV) detects the flow-velocity field and provides data to estimate the shear velocity. In combination with the measured temperature-field, all necessary information for the calculation of the particle deposition and particle relaxation times are available. The measurements are compared to results of theoretical works from the literature. The experimental database is relevant especially for computational fluid dynamics (CFD)-developers, for model development, and model verification. A wide range of phenomena like particle separation, local agglomeration of particles with a specific particle mass, and selective remobilization can be explained in this way. Thus, this work contributes to a realistic analysis of nuclear safety.

Code validation of a scaled-down DRACS model in RELAP5/SCDAPSIM/MOD 4.0

Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system proposed for Fluoride-salt-cooled High-temperature Reactors (FHRs). Benchmark study of decay heat removal capability of DRACS by natural circulation is critical for FHR safety analysis. Two scaled-down test facilities, a low-temperature DRACS test facility (LTDF) and a high-temperature DRACS test facility (HTDF), were designed following a scaling analysis and constructed to investigate thermal performance and heat removal capability of the DRACS. The LTDF uses water as a surrogate coolant for both the primary and DRACS loops. Two transient scenarios were carried out in the LTDF, namely, DRACS startup and primary coolant pump trip. In both of the scenarios, the experimental results demonstrate sufficient heat removal capabilities of the DRACS with natural circulations established in the system. These two scenarios were simulated using RELAP5/SCDASIM/MOD4.0 code and the code simulation results show good agreement with the LTDF experimental data. For the HTDF, molten salts FLiNaK and KF-ZrF4 are used respectively as the primary coolant and DRACS coolant. Thermodynamic and transport properties of FLiNaK and KF-ZrF4 were implemented into RELAP5/SCDAPSIM/MOD 4.0. Simulations of transient scenarios in the HTDF were also performed and the code results will be compared with the HTDF experimental data once available.

Dynamic behavior of a high-temperature printed circuit heat exchanger: Numerical modeling and experimental investigation

Printed circuit heat exchanger (PCHE) is one of the leading intermediate heat exchanger (IHX) candidates for the Very High Temperature Reactors due to its capability for high-temperature and high-pressure applications. An IHX serves to isolate the reactor primary system from electricity generation and process heat application plants, and therefore must be robust to maintain the entire system integrity during normal and off-normal conditions. Understanding the PCHE dynamic behavior is of importance in the design of a PCHE-type IHX since the dynamics of the PCHE will influence accident progression in most accident scenarios. In addition, any transient in the power conversion unit or process heat plant will propagate back to the reactor primary system via the PCHE. In this study, the dynamic behavior of a zigzag-channel PCHE, subject to helium inlet temperature variations and helium mass flow rate step changes on both the hot and cold sides, was simulated and analyzed. Three sets of transient tests were experimentally conducted in a high-temperature helium test facility to examine the dynamic performance of the PCHE and to assess the PCHE dynamic model. Comparisons of the numerical results and experimental data indicate that the numerical solutions are sufficiently accurate and that the feasibility of the dynamic model developed for predicting the steady-state and transient performance of the PCHE is confirmed.

Transient analysis of fluoride-salt-cooled high-temperature reactor test facility based on modified RELAP5/MOD3.2

Transient analysis of fluoride-salt-cooled high-temperature reactor test facility based on modified RELAP5/MOD3.2

Pressure drop and heat transfer characteristics of a high-temperature printed circuit heat exchanger

Printed circuit heat exchanger (PCHE) is one of the leading intermediate heat exchanger (IHX) candidates to be employed in the very-high-temperature gas-cooled reactors (VHTRs) due to its capability for high-temperature, high-pressure applications. In the current study, a reduced-scale zigzag-channel PCHE was fabricated using Alloy 617 plates for the heat exchanger core and Alloy 800H pipes for the headers. The pressure drop and heat transfer characteristics of the PCHE were investigated experimentally in a high-temperature helium test facility (HTHF) at The Ohio State University. The PCHE helium inlet temperatures and pressures were varied up to 464°C/2.7MPa for the cold side and 802°C/2.7MPa for the hot side, respectively, while the maximum helium mass flow rates on both sides of the PCHE reached 39kg/h. The corresponding maximum channel Reynolds number was approximately 3558, covering the laminar flow and laminar-to-turbulent flow transition regimes. New pressure drop and heat transfer correlations for the current zigzag channels with rounded bends were developed based on the experimental data. Comparisons between the experimental data and the results obtained from the available PCHE and straight circular pipe correlations were conducted. Compared to the heat transfer performance in straight circular pipes, the zigzag channels provided little advantage in the laminar flow regime but significant advantage near the transition flow regime.