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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Experimental Investigation of Forced Convection and Natural Circulation Cooling of a VHTR Core under Normal Operation and Accident Scenarios

This NEUP project #15-8205 entitled “Experimental Investigation of Forced Convection and Natural Circulation Cooling of a VHTR Core under Normal Operation and Accident Scenarios”, has investigated important phenomena in the safety of Very High Temperature Reactors (VHTRs) using the experimental facilities at City College of New York (CCNY) and Kansas State University (KSU), in collaboration with the Idaho State University (ISU) and Idaho National Laboratory (INL). The main objectives of this project were to investigate the forced convection and bypass flow phenomena, natural circulation flow and heat transfer, and graphite oxidation due to air ingress which could occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced and natural circulation in a VHTR core. This work has built on the previous NEUP project #11-3218 which investigated forced convection and flow laminarization phenomena in a prismatic core of a VHTR and natural circulation of a pure gas under Pressurized Conduction Cooldown (PCC) and Depressurized Conduction Cooldown (DCC) conditions. Under each Task, extensive literature reviews on natural circulation, bypass flow, air ingress and graphite oxidation phenomena were conducted. In Task 1, scaling analyses were conducted by Idaho State University followed by design and construction of experimental facilities needed to conduct both bypass flow and natural circulation experiments at CCNY. Natural circulation experiments were conducted in Task 2 to investigate the air ingress phenomena, and bypass flow phenomena in Task 3, both at CCNY. In Task 4, radiation heat transfer and graphite oxidation tests were performed at Kansas State University. The experimental data were then analyzed to obtain quantitative results on natural circulation of pure gases and gas mixtures under DCC conditions, bypass flow rates, radiation heat transfer and graphite oxidation. In addition, CFD and Direct Numerical Simulations were conducted by INL and CCNY to investigate the flow laminarization phenomenon in strongly heated gas flows. The scaling analyses have shown that the key nondimensional numbers and parameters all have acceptable distortion levels principally because the CCNY facility’s test sections have representative cooling channel dimensions and are using either helium or a scalable gas such as nitrogen at representative state conditions. The natural circulation experiments involving helium and nitrogen gas mixtures at CCNY have elucidated the effect of nitrogen (simulating air) ingress into a lower plenum on the onset of natural circulation, mass flow rates and heat transfer coefficients under different power input levels in the riser section. Also, the transport of nitrogen from the lower plenum to upper plenum due to natural circulation was quantified for the first time. Nusselt number correlations for the heat transfer coefficient under natural circulation as a function of Reynolds and Prandtl numbers was also developed. The experiments at KSU revealed that prior to the onset of natural circulationin a VHTR core. This work has built on the previous NEUP project #11-3218 which investigated forced convection and flow laminarization phenomena in a prismatic core of a VHTR and natural circulation of a pure gas under Pressurized Conduction Cooldown (PCC) and Depressurized Conduction Cooldown (DCC) conditions. Under each Task, extensive literature reviews on natural circulation, bypass flow, air ingress and graphite oxidation phenomena were conducted. In Task 1, scaling analyses were conducted by Idaho State University followed by design and construction of experimental facilities needed to conduct both bypass flow and natural circulation experiments at CCNY. Natural circulation experiments were conducted in Task 2 to investigate the air ingress phenomena, and bypass flow phenomena in Task 3, both at CCNY. In Task 4, radiation heat transfer and graphite oxidation tests were performed at Kansas State University. The experimental data were then analyzed to obtain quantitative results on natural circulation of pure gases and gas mixtures under DCC conditions, bypass flow rates, radiation heat transfer and graphite oxidation. In addition, CFD and Direct Numerical Simulations were conducted by INL and CCNY to investigate the flow laminarization phenomenon in strongly heated gas flows. The scaling analyses have shown that the key nondimensional numbers and parameters all have acceptable distortion levels principally because the CCNY facility’s test sections have representative cooling channel dimensions and are using either helium or a scalable gas such as nitrogen at representative state conditions.The natural circulation experiments involving helium and nitrogen gas mixtures at CCNY have elucidated the effect of nitrogen (simulating air) ingress into a lower plenum on the onset of natural circulation, mass flow rates and heat transfer coefficients under different power input levels in the riser section. Also, the transport of nitrogen from the lower plenum to upper plenum due to natural circulation was quantified for the first time. Nusselt number correlations for the heat transfer coefficient under natural circulation as a function of Reynolds and Prandtl numbers was also developed. The experiments at KSU revealed that prior to the onset of natural circulation(ONC), molecular diffusion plays a significant role in air-ingress, however, at higher temperatures convection currents may be influencing air ingress and in-turn ONC times. Bypass flow experiments faced difficulties in the measurement of gas velocity in the main channel due to the failure of a high temperature Hot Wire Anemometer probe, so an energy balance method was instead employed to estimate the flow distribution between the main and bypass flow channels. In the graphite oxidation experiments conducted at KSU, oxidation was found to increase the thermal diffusivity of nuclear grade graphite. This change is believed to be driven by oxidation decreasing graphite sample density rather than raising the thermal conductivity. Oxidation was also shown to greatly alter graphite emissivity. If large amounts of impurities have built up within the graphite over time and have significantly increased its effective ash content, then air ingress is likely to severely inhibit the radiative heat transfer abilities of the graphite due to the formation of low emissivity, ash layers on its surface. However, if the purity of the graphite at the time of the air ingress is still close to that of its virgin specifications, the oxidation induced roughening of the graphite’s outer surface would have the effect of slightly raising its emissivity. The flow laminarization phenomenon was investigated by INL by examining the results of Direct Numerical Simulations of previous experiments, and by CCNY conducting its own DNS simulations using DOE’s Nek5000 code

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 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 OSTI2016

R and D activities on VHTR materials at KAERI - 18496

KAERI started VHTR materials R and D in 2006 as part of its role in supporting the development of key technology for the nuclear research and development program. Over the past ten years, research has been conducted on various materials including Alloy 617 for an IHX and hot gas duct, 9Cr-1Mo for an RPV, graphite for a reflector and the support structures in the core region, Cf/C and SiCf/SiC composites for a control rod, and Fe-Si, SiC, Hastelloy{sup R}, and other novel metals for an SI process reactor. Mechanical properties including creep and fatigue, environmental effects, and irradiation behaviors have also been heavily evaluated as part of the materials research. Most of the results have been contributed to the GEN-IV VHTR materials collaboration. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Preliminary structural assessment of VHTR graphite components according to ASME code - 18532

In this study, a preliminary structural assessment on a graphite component of a VHTR was done in accordance with the semi-probabilistic assessment procedures of the latest 2015 edition of the ASME Boiler and Pressure Vessel Code, Section III, Div. 5, Subpart HHA-Graphite Materials. Establishment of the Weibull distribution for the tensile strength is essential to set up the stress limits based on the probability of failure of the graphite materials. All parameters required in the ASME code were identified and evaluated for a nuclear graphite material, IG-110. The assessment of design level was done on one of fuel elements in a Korean VHTR design by two methodologies in the ASME code: (1) Simplified assessment and (2) Full assessment. The structural integrity of the fuel element was assured by both assessment methods. Also, the effect of the number of integration volume groups on the probability of failure in the full assessment was investigated and a recommendation for the conservative results was suggested. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Study on air ingress phenomena during a depressurization accident of a VHTR - 18632

A depressurization accident is the one of the design-basis accidents of a Very-High-Temperature Reactor (VHTR). When a depressurization accident occurs, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Therefore, it is important to understand the mixing processes of different kinds of gases in the stable and unstable stratified fluid layers. In particular, it is also important to examine the influence of localized natural convection and molecular diffusion on the mixing process from a safety viewpoint. Therefore, in order to predict or analyze the air ingress phenomena during a depressurization accident, it is important to develop a method for the prevention of air ingress during an accident. We carried out experiments and numerical analysis using three-dimensional (3D) CFD code to obtain the mixing process of two-component gases and the flow characteristics of localized natural convection. This study also investigated a control method for the natural circulation of air through the injection of helium gas. The numerical model consists of a storage tank and a reverse U-shaped vertical rectangular passage. They are separated by a horizontal partition plate. One sidewall of the high-temperature side vertical passage is heated and the other sidewall is cooled. The low-temperature vertical passage is cooled by ambient air. The experimental procedures and numerical analysis process are as follows. Firstly, the storage tank is filled with heavy gas and the reverse U-shaped vertical passage is filled with a light gas. In the vertical passage of the high-temperature side, localized natural convection is generated by the temperature difference between the vertical walls. These flow characteristics were obtained from the experiments and steady-state analysis. The experiment and analysis began after the partition plate was opened. The results obtained from the experiments were quantitatively simulated using 3D numerical analysis. The two component gases were mixed via molecular diffusion and natural convection. After some time elapsed, natural circulation occurred through the reverse U-shaped vertical passage. These flow characteristics are the same as those of phenomena generated in the passage between a permanent reflector and a pressure vessel wall of the VHTR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Development of TDLAS for Water Vapor Concentration Measurement during VHTR Steam Ingress

Very-high-temperature reactors (VHTR) utilize a graphite-moderated, helium-cooled design that provides higher efficiencies and a low-carbon source of process heat (>700 deg. C) for many industrial applications. Safety concerns have led to extensive accident condition investigations. Steam generator heat-exchanger tube rupture accidents (SGTR) can lead to increased reactor core steam density, which causes neutron slowing-down and chemical corrosion of the graphite fuel elements and structural columns. It is important to be able to computationally model the propagation of steam throughout the lower plenum and nuclear reactor core, but experimental data is needed to validate such models. The primary objective of this project is to instrument a scaled down experiment of a VHTR designed by Ohio State University (OSU), with tunable diode laser absorption spectroscopy (TDLAS) diagnostics to measure steam concentration changes and map how water vapor propagates through the test facility's lower plenum and core. The basic principle of absorption spectroscopy is based on a molecular species ability to absorb photons of specific energies. A collimated laser beam of known energy is passed through a test section containing a sample gas. When a transition of the absorbing gas species resonates with the laser frequency, a certain amount of energy will be absorbed based on the rovibrational characteristics of the molecule. The intensity of the transmitted laser light can be measured by a detector after exiting the test section. The Beer-Lambert Law of absorbance is used to recover this information relating beam attenuation to the system dimensions, gas composition, pressure, and temperature. The line-strength is a function of temperature alone and relates the fraction of molecules in a given rovibrational state through the partition function. The partition function describes the statistical properties of a system in thermodynamic equilibrium where the states obey a Boltzmann distribution. Thus, the number density of molecules in a particular state can be determined from the total internal partition function at a particular temperature. The lineshape function of a given absorption transition is derived from the physical interaction between the absorbing species and is dependent on temperature and pressure. Random thermal motion of absorbing molecules result in broadening of the linewidth, called Doppler broadening, which follows a Gaussian distribution. Molecular collisions result in broadening and shifting of the linewidth, called pressure broadening, which follows a Lorentzian distribution. The overall lineshape results in a Voigt profile, which is a convolution of the Gaussian and Lorentzian distributions. The Voigt profile does not have a simple analytical form and is most often calculated numerically or approximated by complex functions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 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 OSTI2016

Biaxial Thermal Creep of Alloy 617 and Alloy 230 for VHTR Applications

In this study, we employed pressurized creep tubes to investigate the biaxial thermal creep behavior of Inconel 617 (alloy 617) and Haynes 230 (alloy 230). Both alloys are considered to he the primary candidate structural materials for very high-temperature reactors (VITITRs) due to their exceptional high-temperature mechanical properties. The current creep experiments were conducted at 900 degrees C for the effective stress range of 15-35 MPa. For both alloys, complete creep strain development with primary, secondary, and tertiary regimes was observed in all the studied conditions. Tertiary creep was found to he dominant over the entire creep lives of both alloys. With increasing applied creep stress, the fraction of the secondary creep regime decreases. The nucleation, diffusion, and coarsening of creep voids and carbides on grain boundaries were found to be the main reasons for the limited secondary regime and were also found to be the major causes of creep fracture. The creep curves computed using the adjusted creep equation of the form epsilon= cosh 1(1 rt) + P-sigma ntm agree well with the experimental results for both alloys at die temperatures of 850-950 degrees C.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Data Report on the Newest Batch of PCEA Graphite for the VHTR Baseline Graphite Characterization Program

This report details a comparison of mechanical and physical properties from the first billet of extruded PCEA nuclear-grade graphite from the newest batch of PCEA procured from GrafTech. Testing has largely been completed on three of the billets from the original batch of PCEA, with data distributions for those billets exhibiting a much wider range of values when compared to the distributions of properties from other grades. A higher propensity for extremely low values or specimens that broke while machining or handling was also characteristic of the billets from the first batch, owing to unusually large fissures or disparate flaws in the billets in an as-manufactured state. Coordination with GrafTech prior to placing the order for a second batch of PCEA included discussions on these large disparate flaws and how to prevent them during the manufacturing process. This report provides a comparison of the observed data distributions from properties measured in the first billet from the new batch of PCEA with those observed in the original batch, in order that an evaluation of tighter control of the manufacturing process and the outcome of these controls on final properties can be ascertained. Additionally, this billet of PCEA is the first billet to formally include measurements from two alternate test techniques that will become part of the Baseline Graphite Characterization database – the three-point bend test on sub-sized cylinders and the Brazilian disc splitting tensile strength test. As the program moves forward, property distributions from these two tests will be based on specimen geometries that match specimen geometries being used in the irradiated Advanced Graphite Creep (AGC) program. This will allow a more thorough evaluation of both the utility of the test and expected variability in properties when using those approaches on the constrained geometries of specimens irradiated in the Advanced Test Reactor as part of the AGC experiment.

36 MATERIALS SCIENCE↗