Bypass Gaps Research Articles

CFD Investigation of MHTGR Natural Circulation and Decay Heat Removal in P-LOFC Accident

One of the key postulated accidents in a high-temperature gas-cooled reactor (HTGR) is the pressurized loss of forced circulation (P-LOFC) of the primary loop, which can be triggered by its primary helium circulator trip or turbine trip. If the reactor shutdown cooling system (SCS) fails during a P-LOFC accident, part of the reactor decay heat is absorbed by the reactor core materials and the rest removed by the reactor cavity cooling system (RCCS). In the extended period of P-LOFC accident compound with SCS failure, the core decay heat is supposed to be removed by conduction, natural circulation, convection, and radiation. A three-dimensional (3-D) computational fluid dynamics (CFD) simulation was performed in this research to study the long-term heat removal mechanisms in the General Atomics’ Modular High Temperature Gas-cooled Reactor (MHTGR) design in a P-LOFC accident. The reactor core temperature distribution and flow field were obtained at different decay power levels. The sensitivity of the natural circulation flow to the bypass gap width was investigated. The natural circulation flow intensity is relatively weak but very sensitive to the width of the bypass gaps.

Uniform Heated Scaled-Down Standard Fuel Block Test to Validate Core Thermofluid Analysis Code for Prismatic Gas-Cooled Reactor

The Korea Atomic Energy Research Institute (KAERI) has developed the Core Reliable Optimization and thermofluid Network Analysis (CORONA) code for core thermofluid analysis of a prismatic high-temperature gas-cooled reactor (HTGR). KAERI performed scaled-down standard fuel block (SFB) heated tests at a helium experimental loop to validate the CORONA code. The scaled-down SFB was designed based on the core thermofluid design for a 350-MW(thermal) HTGR. The reference test condition was selected to maintain the Reynolds number of the coolant channels and the bypass gaps. The test section had seven coolant holes and 12 fuel holes considering KAERI’s helium loop circulator design. The material of the fuel block was Al2O3, selected to simulate the low thermal conductivity of the irradiated graphite at the high-temperature condition. The bypass gap structure was made of stainless steel 304 to minimize gap size deformation at the heated condition. This paper presents a comparison between the test results and the CORONA analysis results. The test parameter was the nitrogen flow velocity (3.6 to 6.0 kg/min) and constant heated condition.

Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200

The core of the very high temperature reactor (VHTR) PMR200 (a prismatic modular reactor rated at 200MW of thermal power) consists of hexagonal prismatic fuel blocks and reflector blocks made of graphite. If the core bypass flow ratio increases, the coolant channel flow is decreased and can then lower the heat removal efficiency, resulting in a locally increased fuel block temperature. The coolant channels in the fuel blocks are connected to bypass gaps by the cross gap, complicating flow distribution in the VHTR core. Therefore, reliable estimation of the bypass flow is highly important for the design and safety analysis of the VHTR core. Because of the complexity of the core geometry and gap configuration, it is challenging to predict the flow distribution in the VHTR core. To analyze this flow distribution accurately, it is necessary to determine the cross flow phenomena, and the loss coefficient across the cross gap has to be evaluated to determine the flow distribution in the VHTR core when a lumped parameter code or a flow network analysis code that uses the correlation of the loss coefficient is employed. The purpose of this paper is to develop a loss coefficient correlation applicable to the cross gap in the PMR200 core. The cross flow was evaluated experimentally using the difference between the measured inlet and outlet mass flow rates. Next, the applicability of a commercial computational fluid dynamics (CFD) code, CFX 15, was confirmed by comparing the experimental data and CFD analysis results. To understand the cross flow phenomena, the loss coefficient was evaluated; in the high Reynolds number region, the cross flow loss coefficient on Reynolds number is nearly constant regardless of the Reynolds number, whereas it varies with the gap size in the low Reynolds number region. Finally, the loss coefficient correlation was proposed on the basis of experiments and the CFD analysis. The developed correlation was compared with existing correlations, and the developed one shows better agreement with the experimental results than the existing ones. Hence, the developed correlation will be applied to the flow network code to analyze flow distribution in the PMR200 core.

Fuel temperature prediction using a variable bypass gap size in the prismatic VHTR

The temperature gradient and hot spot temperatures were calculated in the prismatic very high temperature reactor as a function of the variable bypass gap size. Many previous studies have predicted the temperature of the reactor core based on a fixed bypass gap size. The graphite matrix of the assemblies in the reactor core undergoes a dimensional change during the operation due to thermal expansion and neutron fluence. The expansion and shrinkage of the bypass gaps change the coolant flow fractions into the coolant channels, the control rod holes, and the bypass gaps. Therefore, the temperature of the assemblies may differ compared to those for the fixed bypass gap case. The temperature gradient and the hot spot temperatures are important for the design of reactor structures to ensure their safety and efficiency. In the present study, the temperature variation of the PMR200 is studied at the beginning (BOC), middle (MOC), and end (EOC) of cycles 07 and 14. CORONA code which has been developed in KAERI is applied to solve the thermal-hydraulics of the reactor core of the PMR200. CORONA solves a fluid region using a one-dimensional formulation and a solid region using a three-dimensional formulation to enhance the computational speed and still obtain a reasonable accuracy. The maximum temperatures in the fuel assemblies using the variable bypass gaps did not differ much from the corresponding temperatures using the fixed bypass gaps. However, the maximum temperatures in the reflector assemblies using the variable bypass gaps differ significantly from the corresponding temperatures using the fixed bypass gaps.

A Steady-State Thermal-Hydraulic Analysis Method for Prismatic High-Temperature Gas-Cooled Reactors

A new methodology for the accurate and efficient determination of steady-state thermal-hydraulic parameters for prismatic high-temperature gas reactors is developed. Whole-core steady-state temperature, pressure, and mass flow distributions are determined for the conceptual MHTGR-350 [Modular High Temperature Gas Reactor] reactor design and also for a range of values of the important parameters. Full-core three-dimensional heat conduction calculations are performed at the individual fuel pin and lattice assembly block levels. A simplified one-dimensional fluid model is developed to predict convective heat removal rates from solid core nodes. Downstream fluid properties are determined by performing a channel energy balance along the axial node length. To establish flow distribution, channel exit pressures are compared, and inlet mass flows are adjusted until a uniform outlet pressure is reached. Bypass gaps between assembly blocks as well as coolant channels are modeled. Finite volume discretization of energy and momentum conservation equations are formulated and explicitly integrated in time. Iterations are performed until all local core temperatures stabilize and global convective heat removal matches heat generation.Whole-core steady-state, thermal-hydraulic results are presented for various axial power and uniform radial power configurations. For all cases, peak temperatures were below expected normal operational limits for TRISO fuels. Bottom-peaked axial power shapes had the highest peak temperatures but the lowest average temperatures. Different reactor designs with increased core inlet temperatures, reduced flow rates, or higher-power-density fuels could however challenge temperature limits. Partial assembly hydrodynamic and temperature results compared favorably with those available in the literature for similar analyses.

Experimental study of core bypass flow in a prismatic VHTR based on a two-layer block model

Bypass flow in a prismatic very high temperature gas-cooled nuclear reactor (VHTR) plays an important role in determining the coolant distribution in the core region. Efficient removal of heat from the core relies on the majority of coolant passing through the coolant channels instead of the bypass gaps. Consequently, the bypass flow fraction and its flow characteristic are important in the design process of the prismatic VHTR. The objective of this study is to experimentally investigate the flow behavior including the turbulence characteristics inside the bypass gaps using laser Doppler velocimetry (LDV), bypass fraction and pressure drops in the system. The experiment facility constructed at Texas A&M University is a scaled model consisting of two layers of fuel blocks. The distributions of the mean streamwise velocity, turbulence intensity and turbulence kinetic energy within the bypass gap at two different elevations under different Reynolds number were investigated. Uncertainties in the bypass flow fraction estimation were evaluated. The velocity and turbulence study in this work is considered to be unique, and may serve as a benchmark for the related numerical calculations.

Study on Influence of Turbulence Model Selection on Prediction of Flow Distribution and Hot Spot Fuel Temperature in Prismatic HTGR Cores

CFD analyses were performed for a cross-flow experiment in prismatic HTGR core fuel blocks to investigate the effect of turbulence models on the prediction of the core flow distribution. A transition model was considered in addition to a fully turbulent model because the flow in the coolant channels of the fuel block is turbulent whereas that in the bypass gap between the blocks is laminar. Additional analyses were carried out to investigate how the variation of the coolant channel flow dependence on the turbulence model selection affected the hot spot fuel temperature. The results showed that the gamma-theta transition model led to a better prediction than the others because it reasonably simulated the flow region where the bypass and cross gaps intersected. The result for the hot spot fuel temperature, however, showed that the temperature change caused by the turbulence model selection is so small that it can be absorbed in a margin of the engineering design. Therefore, the fully turbulent model can be used as an efficient alternative to the transition model in the engineering calculation of an HTGR core flow.

Bypass flow computations on the LOFA transient in a VHTR

Bypass flow in the prismatic gas-cooled very high temperature reactor (VHTR) is not intentionally designed to occur, but is present in the gaps between graphite blocks. Previous studies of the bypass flow in the core indicated that the cooling provided by flow in the bypass gaps had a significant effect on temperature and flow distributions for normal operating conditions. However, the flow and heat transports in the core are changed significantly after a Loss of Flow Accident (LOFA). This study aims to study the effect and role of the bypass flow after a LOFA in terms of the temperature and flow distributions and for the heat transport out of the core by natural convection of the coolant for a 1/12 symmetric section of the active core which is composed of images and mirror images of two sub-region models. The two sub-region models, 9 × 1/12 and 15 × 1/12 symmetric sectors of the active core, are employed as the CFD flow models using computational grid systems of 70.2 million and 117 million nodes, respectively. It is concluded that the effect of bypass flow is significant for the initial conditions and the beginning of LOFA, but the bypass flow has little effect after a long period of time in the transient computation of natural circulation.

Development of a thermo-hydraulic bypass leakage test method for the Wendelstein 7-X target element cooling structure

A facility for testing the cooling structure to ensure the quality of the target element heat sink is under construction at IPP Garching. This test bed has been built up to do a proof of concept study with a hydraulic mock up, built from the same material as the target elements, but can be opened for artificial gap manufacture. A bypass in the cooling structure was the main reason for the overheating of CFC tiles, which resulted in a defect by cracking of the CFC–AMC-Cu interlayer in pre-series 3. The cooling structure is built from two half-shells, in both a half pipe is milled [1]. To correlate the thermal response function to the artificial bypass gaps the time constant of the cool down was used. The results of the measurements are presented and compared with the calculated results.

Velocity distribution and vibration excitation in tube bundle heat exchangers

Design criteria for tube bundle heat exchangers, to avoid fluidelastic instability, are based on stability criteria for ideal bundles and uniform flow conditions along the tube length. In real heat exchangers, a non-uniform flow distribution is caused by inlet nozzles, impingement plates, baffles and bypass gaps. The calculation of the equivalent velocities, according to the extended stability equation of Connors, requires the knowledge of the mode shape and the assumption of a realistic velocity distribution in each flow section of the heat exchanger. It is the object of this investigation to derive simple correlations and recommendations, (1) for equivalent velocity distributions, based on partial constant velocities, and (2) for the calculation of the critical volume flow in practical design applications. With computational fluid dynamic (CFD) programs it is possible to calculate the velocity distribution in real tube bundles, and to determine the most endangered tube and thereby the critical volume flow. The paper moreover presents results and design equations for the inlet section of heat exchangers with variations of a broad range of geometrical parameters, e.g., tube pitch, shell diameter, nozzle diameter, span width, distance between nozzle exit and tube bundle.