Very High Temperature Gas-cooled Reactor Research Articles

High-temperature mechanical behaviors of diffusion-welded Alloy 617

The development of a compact-type heat exchanger is one of the critical issues for realizing an efficient gas-to-gas heat exchanging system in a very high-temperature gas-cooled reactor (VHTR). The nickel-based superalloy Alloy 617 (Ni-Cr-Co-Mo, UNS N06617) is considered as the leading material for this application owing to the excellent thermal stability and strength in high temperature ranges. Sixty sheets having dimensions of 200 × 200 mm2 were diffusion-welded under uniaxial compressive pressure of 14.7 MPa at 1423 K (1150 °C) in a high vacuum condition. After 2 h of hot pressing, a post-weld heat treatment (PWHT) was applied to enhance the constituent atomic diffusion across the interface. To evaluate the high-temperature mechanical properties, tensile and stress-rupture tests were employed in this study. In tensile tests, the yield strength of the diffusion weldment is 318 ± 4 MPa at room temperature, and superior yield strengths over the ASME code requirement are maintained up to 1223 K (950 °C). Similarly, the tensile strength reaches 730 ± 15 MPa with ductility of 37.1 ± 1.4% at room temperature. However, both the tensile strength and ductility drop rapidly at elevated temperatures (≥973 K (700 °C)). During stress-rupture testing at a fixed temperature of 1173 K (900 °C) in an ambient air environment, the diffusion weldment shows lower creep properties. A plot of log time-to-rupture vs. log minimum creep rate is illustrated on the basis of the Monkman-Grant (M–G) relationship.

Time-dependent neutron diffusion analysis using finite element method for a block-type VHTR core design

The very high-temperature gas-cooled reactor (VHTR) has attracted interest owing to its high passive safety. A reactor analysis tool is required to analyze the safety characteristics of a VHTR in detail, and such a tool should be capable of a transient calculation for a reactivity insertion accident such as control rod ejection or withdrawal. This work presents a transient analysis of a block-type VHTR core. Time-dependent neutron diffusion equation is solved by the finite element method. A simplified thermal fluid analysis tool is also implemented to consider thermal feedback. In addition, a new method is introduced to resolve the control rod cusping effect without additional computation or mesh reconstruction. The above methods are applied to a reactor physics code CAPP, developed at Korea Atomic Energy Research Institute (KAERI). The methods are tested on three-dimensional VHTR problems that include control rods and show encouraging results.

Combined cycle-coupled high-temperature and very high-temperature gas-cooled reactors: Part I—Cycle optimization

Very high-temperature gas-cooled reactors (VHTRs) have attracted worldwide attention because of their multiple uses, which include hydrogen production and high-efficiency power generation. Efficient power conversion is a key to improving the competitiveness of VHTRs. The combined cycle is a cyclic form of energy cascade utilization that makes full use of the high-temperature heat source of VHTRs. In this paper, a theoretical optimization method of combined-cycle efficiency is proposed for combined cycle-coupled high-temperature gas-cooled reactors (HTGRs) and VHTRs without considering specific engineering constraints. Based on a total differential analysis combined with a numerical analysis, the multi-variable combined-cycle efficiency was reduced to a function of two key parameters: the reactor outlet temperature and the main steam pressure. The remaining parameters can be determined by optimization within the defined domain. The theoretical optimization method proposed in this paper is conceptually clear and solves the problems presented by the many variables involved in combined cycle-coupled HTGRs without a clear variable optimization strategy. The results of this study provide guidance for the theoretical analysis and parameter design of combined cycle-coupled VHTRs.

Three-dimensional looped network analysis code including core thermal analysis model for prismatic very high temperature gas-cooled reactor

The core of a prismatic very high temperature gas-cooled reactor (VHTR) is composed of stacked graphite blocks with gaps between them, which results in undesired flows through the gaps. These flows complicate the flow distribution in the reactor core and cause difficulty in predicting the temperature distribution of the graphite block. Conventionally, computational fluid dynamics (CFD) codes have been mainly used for the VHTR reactor core analysis. However, they require considerable calculation time and cost, and, therefore, are considered too expensive in terms of calculation time to investigate the effect of the gap size distribution in the core. As numerous cases with different gap size combinations need to be tested in reactor design, it can be said that high calculation speed of the design code with reasonable accuracy is an important feature.In this study, a thermo-fluid analysis code for the core of a prismatic VHTR, named FastNet (Flow Analysis for Steady-state Network), was developed for prediction of the core flow and temperature distribution with affordable computational cost. For rapid calculation, a flow network analysis method was used for flow distribution analysis, and a thermal analysis model was added to analyze the whole core temperature distribution. To overcome the drawbacks of its low resolution, an effective thermal conductivity model and a maximum fuel temperature model were applied. Finally, to verify the code, results of the FastNet calculation were compared to other codes such as the CFD code and CORONA code as a code-to-code validation. The results show that a satisfactory accuracy was obtained with a remarkably short computational time.

Verification of DeCART2D/CAPP code system for VHTR analysis with PMR-200 benchmark

This paper presents the verification of the DeCART2D/CAPP code system for the Very High Temperature Gas-Cooled Reactor (VHTR) analysis with the Prismatic Modular Reactor 200 (PMR-200) benchmark. The McCARD Monte Carlo (MC) code is used to obtain the reference solution. The verification has been performed for the effective multiplication factor (keff) and reactivity coefficients at the levels of fuel compact, fuel block, and full core. Furthermore, the verification of the depletion calculation has been conducted for the fuel block and the verification for the power distribution has been performed at the levels of fuel block, two-dimensional (2D) and three-dimensional (3D) full core. The verification results of DeCART2D, CAPP, and DeCART2D/CAPP are compared systematically against the reference McCARD solutions to demonstrate the VHTR modeling capability and accuracy of the codes. It was successfully shown that the keff errors of the DeCART2D/CAPP code system are smaller than ∼510 pcm, the isothermal temperature coefficient (ITC) errors are smaller than ∼0.66 pcm/K, and the power distribution errors are smaller than 2.80%. It was also shown that the maximum keff errors of DeCART2D fuel block depletion calculations are smaller than ∼460 pcm.

Flow visualization experiment in a two-side wall heated rectangular duct for turbulence model assessment in natural convection heat transfer

Abstract Reactor Cavity Cooling System (RCCS) is one of the passive safety systems in Very High Temperature gas-cooled Reactor (VHTR). RCCS incorporates riser channels connected with a chimney to remove the residual heat emitted from the reactor vessel using natural circulation, and therefore, heat transfer phenomena in the RCCS riser duct is of great importance to ensure the safety of the VHTR. To enhance the understanding of heat transfer mechanism in a heated riser duct with rectangular shape, visualization of the local flow structure was conducted with an experimental facility for the natural convection heat transfer applying the Particle Image Velocimetry (PIV) method. With measured thermal boundary conditions by IR thermometry, CFD analysis was performed for the test facility using two different turbulence models. By comparing local flow structure from the experimental data and CFD calculation results, turbulence model assessment was conducted to confirm the prediction capability of the models. Experimental data show the magnitudes and locations of maximum velocity peak which is induced by buoyancy force and the anisotropic behaviors of velocity fluctuations near wall region. CFD calculations overestimate the maximum velocity near the wall compared with experimental data, and they seem to have limitations in reproducing the secondary flow observed at the corner of the test section in the experiment. The established methodology for flow visualization and the measurement strategy for thermal boundary conditions could contribute to obtain extensive experimental data and turbulence model assessment results, which would improve the understanding of the heat transfer mechanism to enhance the performance prediction capability for the passive heat removal system of the VHTR.

VHTR core analysis with McCARD and DeCART codes for high temperature engineering test reactor benchmark

This paper presents a Very High Temperature Gas-Cooled Reactor (VHTR) core analysis of the McCARD and DeCART-2D with High Temperature Engineering Test Reactor (HTTR) benchmark problem. The numerical results of the HTTR benchmark problem are demonstrated, the capability for VHTR core analysis of McCARD is validated by using the experiment data, and that of DeCART-2D is verified by comparing the numerical results with those of McCARD. The detail HTTR benchmark problem and specification of core geometry and material are also described. Also, a method of approximation for the complex geometry of the VHTR core is described for modelling in computational code. In terms of the multiplication factor and temperature coefficient, the numerical results of McCARD are easily compared with the HTTR experiment data. Because the DeCART-2D can solve 2-dimensional (2D) geometry, the simulation of 2D HTTR fuel rod, block, and core problem is performed with DeCART-2D and McCARD. The numerical results between them are around 200–400 pcm, meaning that DeCART-2D can be used to analyze VHTR core.

Transient heat transfer for helium gas flowing over a horizontal cylinder in a narrow channel

ABSTRACTTransient forced convection heat transfer due to exponentially increasing heat input to a heater is important as a database for the safety assessment of the transient heat transfer process in a very high temperature gas-cooled reactor (VHTR). Forced convection transient heat transfer for helium gas flowing over a cylinder in a narrow channel was experimentally studied at various periods of exponential increases in heat input. The test heater was mounted horizontally along the center part of a circular test channel with a diameter of 5 mm. The heat generation rates of the heater,, was increased with the exponential function, . By using a narrow channel, relatively high flow velocity was achieved, and experimental data at a high Reynolds number were obtained. According to the results, the surface temperature and the heat flux exponentially increased with time. It was clarified that the heat transfer coefficient approached the quasi-steady-state for a period of more than approximately 1 s, and it reached higher values for a period of less than approximately 1 s. The heat transfer coefficients present high dependence on the flow velocity of the helium gas and the heater diameter. Heat transfer correlations at the quasi-steady state and the transient state were obtained based on the experimental data.

Thermal hydrodynamic modeling and simulation of hot-gas duct for next-generation nuclear reactor

A very high-temperature gas-cooled reactor (VHTR) is a fourth-generation nuclear power reactor that requires an intermediate loop that consists of a hot-gas duct (HGD), an intermediate heat exchanger (IHX), and a process heat exchanger for massive hydrogen production. In this study, a mathematical model and simulation were developed for the HGD in a small-scale nitrogen gas loop that was designed and manufactured by the Korea Atomic Energy Research Institute. These were used to investigate the effect of various important factors on the surface of the HGD. In the modeling, a porous model was considered for a Kaowool insulator inside the HGD. The natural convection and radiation are included in the model. For validation, the modeled external surface temperatures are compared with experimental results obtained while changing the inlet temperatures of the nitrogen working fluid. The simulation results show very good agreement with the experiments. The external surface temperatures of the HGD are obtained with respect to the porosity of insulator, emissivity of radiation, and pressure of the working fluid. The local Nusselt number is also obtained for the optimum design of a possible next-generation HGD.

Research on the High-Temperature Hot Compressive Deformation Behavior of Ni-Based Superalloy GH3128

Intermediate heat exchanger (IHX), which transfers the heat generated in the reactor core to the secondary loop, is one of the key structural components of the very high-temperature gas-cooled reactor (VHTR). The Ni-based superalloy GH3128 has good high-temperature strength and so is a promising main structural material for the IHX. In this paper, the flow stress behaviors and the deformation microstructure of superalloy GH3128 were investigated by high-temperature compression tests conducted at various temperatures (950–1150 °C) and strain rates (0.001–10 s−1), and the processing maps were analyzed in order to establish the hot deformation constitutive model and obtain the optimum hot forming condition. The results show that (1) both flow stresses and peak flow stresses increase along with the increase of strain rate or decrease of temperature, (2) GH3128 has excellent hot workability, (3) the dynamic recovery (DRV) plays the dominant role during the dynamic softening process due to the high stack fault energy, and (4) the optimum hot forming condition of GH3128 should be defined in the temperature of 1150 °C and strain rate range of 0.01–0.056 s−1. This work contributes to the application of GH3128 alloy on IHX structure.

Numerical solution of heat transfer process in a prismatic VHTR core accompanying bypass and cross flows

Bypass flow and cross flow gaps both exist in the core of a very high temperature gas-cooled reactor (VHTR), which is inevitable owing to tolerances in manufacturing, thermal expansion and irradiation shrinkage. The coolant mass flow rate distribution, temperature distribution, and hot spot temperature are significantly affected by bypass and cross flows. In the present study, three-dimensional CFD analysis is conducted for thermal analysis of the reactor core. A validation study for the turbulence model is performed by comparing the friction coefficient with published correlations. A sensitivity study of the near wall mesh is conducted to ensure mesh quality. Parametric studies are performed by changing the size of the bypass and cross gaps using a one-twelfth sector of a fuel block. Simulation results show the influence of the bypass gap size on temperature distribution and coolant mass flow rate distribution in the prismatic core. It is shown that the maximum fuel and coolant channel outlet temperatures increase with an increase in the gap size, which may lead to a structural risk to the fuel block. The cross flow is divided into two types: the cross flow from the bypass gap to the coolant channels and the cross flow from the high-pressure coolant channels to low-pressure coolant channels. These two types of flow have an opposing influence on the temperature gradient. It is found that the presence of the cross flow gaps may have a significant effect on the distribution of the coolant in the core due to flow mixing in the cross gaps.

Experimental and numerical study of a printed circuit heat exchanger

Printed circuit heat exchangers (PCHEs) are promising to be employed in very-high-temperature gas-cooled reactors (VHTRs) due to their high robustness for high-temperature, high-pressure applications and high compactness. PCHEs typically serve as intermediate heat exchangers (IHXs) that isolate the secondary loop from the reactor’s primary system and hence must be sufficiently robust to maintain the system integrity during normal and off-normal conditions. In addition, the performance of the PCHE-type IHX could considerably affect the nuclear power plant overall operation since any transients on the secondary side would be propagated back to the reactor’s primary coolant system via the IHX. It is therefore imperative to understand how the PCHE would dynamically respond to a variety of transients. In the current study, experiments were first conducted to examine the steady-state thermal performance of a reduced-scale straight-channel PCHE. A dynamic model benchmarked in a previous study was then used to predict the steady-state and transient behavior of the PCHE. The steady-state temperature profiles of the working fluids on both the hot and cold sides and in the solid plates of the heat exchanger were obtained, which served as the initial condition for the transient simulations. The detailed dynamic response of the straight-channel PCHE, subject to inlet temperature variations, helium mass flow variations, and combinations of the two, was simulated and analyzed. In addition, two sets of transient tests, one for helium inlet temperature increase and the other for helium inlet temperature decrease, were experimentally carried out to assess the applicability of the dynamic model. Comparisons of the numerical results with the experimental data show that the dynamic model is successful in predicting the experimental transient scenarios. Although difference was observed between the numerical results and experimental data, the comparisons suggest that the numerical solutions are sufficiently accurate and conservative and that the applicability of the dynamic model proposed for predicting the steady-state and transient performance of the straight-channel PCHE has been confirmed. Furthermore, both the numerical and experimental studies provide insights into the dynamic performance of the PCHE.

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.

Investigation of absorption characteristics for thermal-load fluctuation using HTTR

ABSTRACTA commercial very high-temperature gas-cooled reactor (VHTR) hydrogen cogeneration system named gas turbine high temperature reactor 300-cogeneration (GTHTR300C) is designed and developed in Japan Atomic Energy Agency (JAEA). Moreover, it has been planned that hydrogen production system and gas turbine system is connect to high-temperature engineering test reactor (HTTR). The reactor system is required to continue a stable and safety operation as well as a stable power supply in the case that thermal-load is fluctuated by the occurrence of abnormal event in the heat utilization system like as the hydrogen production facility. Then, it is necessary to confirm that the thermal-load fluctuation could be absorbed by the reactor system so as to continue the stable and safety operation. The thermal-load fluctuation absorption tests using the HTTR were planned to clarify the absorption characteristics of the HTGR system. However, it is difficult to clarify the phenomenon due to many kinds of fluctuation in nuclear thermal power in the reactor core. Moreover, the actual data regarding how the delay of the temperature response is effective for the reactor system had been gained quantitatively.The thermal-load fluctuation absorption tests without nuclear heating were planned and conducted in JAEA to clarify the absorption characteristic of thermal-load fluctuation mainly by the reactor and by the intermediate heat exchanger (IHX). The absorption characteristics of thermal-load fluctuation can be revealed with sufficient temperature fluctuation. So the tests were conducted with the primary coolant temperature 120 °C which is the start-up temperature of the HTTR. As a result it was revealed that the reactor has the larger absorption capacity of thermal-load fluctuation than the expected one and that the IHX can be contributed to the absorption of the thermal-load fluctuation generated in the heat utilization system in the reactor system. It was confirmed from their results that the reactor and the IHX has effective absorption capacity of the thermal-load fluctuation generated in the heat utilization system. Moreover, the calculation with the safety evaluation code based on RELAP5/MOD3 was performed. It was confirmed that the calculated temperature for the reactor is almost same to the measured one with the new analysis model. On the other hand, it was confirmed that the calculated temperature for the IHX decreased faster than the measured one due to smaller absorption capacity in the calculation model than that in actual one. It can be considered that the calculation for the IHX produces the conservative result. It was summarized that the safety evaluation code can represent the thermal-load fluctuation absorption behavior conservatively.

Computational fluid dynamics analysis of the initial stages of a VHTR air-ingress accident using a scaled-down model

An air-ingress accident is considered to be one of the design basis accidents of a very high-temperature gas-cooled reactor (VHTR). The air-ingress accident is initiated, in its worst-case scenario, by a complete break of the hot duct in what is referred to as a double-ended guillotine break. This leads to an initial loss of the primary helium coolant via depressurization. Following the depressurization process, the air–helium mixture in the reactor cavity could enter the reactor core via the hot duct and hot exit plenum. In the event that air ingresses into the reactor vessel, the high-temperature graphite structures in the reactor core and hot plenum will chemically react with the air, which could lead to damage of in-core graphite structures and fuel, release of carbon monoxide and carbon dioxide, core heat up, failure of the structural integrity of the system, and eventually the release of radionuclides to the environment.Studies in the available literature focus on the phenomena of the air ingress accident that occur after the termination of the depressurization, such as density-driven stratified flow, molecular diffusion, and natural circulation. However, a recent study shows that flow reversal could occur due to Taylor wave expansion near the end of the depressurization, which could affect subsequent stages of the air ingress accident scenario. Therefore, to properly understand and evaluate the depressurization effects, numerical simulations are performed for the double-ended guillotine break of the Gas Turbine-Modular Helium Reactor (GT-MHR) cross vessel with a computational fluid dynamics (CFD) tool, ANSYS FLUENT.A benchmark and error quantification study of the depressurization shows that the ANSYS FLUENT model can predict the depressurization problem with relatively low uncertainty. In addition, the computational results show that the depressurization of a double-ended guillotine break behaves as an isentropic process. The observed flow oscillations near the end of the depressurization promote mixing of helium gas and air near the break. The results of the CFD analyses also show that the density-driven stratified flow, which is postulated to be the next stage of the air-ingress accident scenario, is strongly dependent on the density difference between the air–helium mixture in the containment and the helium in the reactor vessel. Therefore, the flow oscillations near the end of the depressurization stage may have a minor, yet notable, effect to slow down the air ingress due to density-driven stratified flow by decreasing the bulk density of the gas mixture in the containment through the addition of helium and increasing the bulk density in the reactor vessel through the addition of air.

Materials for Nuclear and Fossil Energy Applications

We are extremely happy to release this special issue. Energy is the prime driver for modern civilization. Strong inverse correlation between per capita energy consumption and poverty level suggests that, in order to alleviate poverty and enhancement of general living standards, the global energy consumptionwill growmanyfold in the coming decades.With presentday technologies, this is simply not feasible particularly for the choice of present-day resource portfolio and the associated economic cost of scaling. For example, according to one estimate, for the global population to achieve western European living standard, characterized by moderate rate of per capita energy consumption (≈7 kW compared to ≈11 kW for USA), we will need $2.2 trillion per year investment for next 20 years in a sustained rate at present-day prices. While one can debate about these numbers, what is clear is that, without technological advancements and breakthroughs, majority of the global population will have no choice but to settle for very poor living standards. Global climate change and deteriorating air quality in urban areas are also a direct consequence of how we source and utilize energy. Since the growth rate of cleaner energy sources is far lower than needed, serious technical research is necessary for better utilization of the traditional resources. Generally, materials are the bottleneck in developing new methods and technologies and hence, in this issue, we have tried to encompass a broad range of topics related to the materials aspects of nuclear energy-based and fossil fuel-based energy generation. Articles presented in this issue are related to Sodiumcooled Fast Reactors (SFRs), Very High Temperature gascooled Reactors (VHTRs), Pressurized Water Reactors (PWRs), and other reactors. Alloy development for nuclear fuel rods is an important topic in nuclear technology, and in this issue the article by J.-H. Kim and coworkers discusses the development of an injection casting process for preparing fuel slugs of U-10Zr-5Mn, a surrogate alloy for use in Sodiumcooled Fast Reactors (SFRs). Alloys to be used in energy generation plants have stringent requirements in terms of high temperature mechanical properties. In this issue, S. J. Kim and coworkers have presented a comparison of cyclic stress response and fracture behavior of Alloy 617 in the form of base metal and in the form of welded joints. Another critical issue in nuclear power generation is corrosion inhibition of the inner linings of the pipes and the tubes. These alloys get corroded due to exposure to primary water of nuclear power plants. Surface treatment of the alloys and/or inhibitor addition to the primary water are two approaches to inhibit corrosion, and these approaches are discussed in the articles by K. M. Kim et al. and by S. J. Kim et al., respectively. Materials challenges in nuclear wastemanagement and in recovery of nuclear fuel from spent fuel rods have also been considered in this issue. In addition, design of new generation of low power thorium-based nuclear technology also forms a part of this issue.

Thermal Predictions of the AGR-3/4 Experiment With Time-Varying Gas Gaps

A thermal analysis was performed for the Advanced Gas Reactor test experiment (AGR-3/4) with time-varying gas gaps. The experiment was irradiated at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Several fuel irradiation experiments are planned for the AGR Fuel Development and Qualification Program that supports the development of the Very-High-Temperature Gas-Cooled Reactor (VHTR) under the Next-Generation Nuclear Plant (NGNP) project. AGR-3/4 combines two tests in a series of planned AGR experiments to test tristructural-isotropic (TRISO)-coated, low-enriched uranium oxy-carbide fuel. Forty-eight TRISO-fueled compacts were inserted into 12 separate capsules for the experiment (four compacts per capsule). The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures using time (fast neutron fluence)-varying gas gaps and to compare with experimentally measured thermocouple data. A finite-element heat transfer model was created for each capsule using the commercial code ABAQUS. Model results are compared to thermocouple data taken during the experiment.

Analytical Study on the Effective Thermal Conductivity of VHTR Fuel Block Geometry with Multiple Cylindrical Holes

In Korea, the Very High Temperature Gas-Cooled Reactor (VHTR) PMR200 is being developed in the Nuclear Hydrogen Development and Demonstration project. Its core consists of hexagonal prism-shaped graphite blocks for the fuel and reflector, and each hexagonal fuel block contains 108 cylindrical coolant holes and 210 fuel compacts. Because of these holes and fuels, the heat transfer in lateral directions in the fuel blocks becomes very complicated. Especially in accident situations when forced convection is lost, the majority of the afterheat flows in the radial direction by conduction across the large number of coolant holes. Moreover, radiation heat transfer is supposed to be added to the radial heat transfer modes owing to the high temperature of the VHTR core. Because of these complexities in radial heat transfer, reliable modeling for effective thermal conductivity (ETC) is required in order to analyze the reactor core thermal behavior using lumped-parameter codes, which are often used to evaluate the integrity of nuclear fuel embedded in the graphite block. In this study, the ETC model adopted in the GAMMA+ code was introduced, and the adequacy of the model was assessed by the commercial computational fluid dynamics (CFD) code CFX-13. The results of the CFD analysis were consistent with the ETC model in general even if a slight disagreement was shown for the case of high temperature. From these analyses, it could be concluded that the ETC model adopted in the GAMMA+ code is an adequate model for the analysis of the PMR200 reactor core. Moreover, it was found that the effect of fuel gap can cause an overprediction of the ETC if the fuel compact thermal conductivity is larger than the applicable range of the model.

Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

A critical event in the safety analysis of the very high-temperature gas-cooled reactor (VHTR) is an air-ingress accident. This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the coaxial cross vessel, which leads to a rapid reactor vessel depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. Following the vessel depressurization, the dominant mode of ingress of an air–helium mixture into the reactor vessel will either be molecular diffusion or density-driven stratified flow. The mode of ingress is hypothesized to depend largely on the break conditions of the cross vessel. Since the time scales of these two ingress phenomena differ by orders of magnitude, it is imperative to understand under which conditions each of these mechanisms will dominate in the air ingress process.Computer models have been developed to analyze this type of accident scenario. There are, however, limited experimental data available to understand the phenomenology of the air-ingress accident and to validate these models. Therefore, there is a need to design and construct a scaled-down experimental test facility to simulate the air-ingress accident scenarios and to collect experimental data. The current paper focuses on the analyses performed for the design and operation of a 1/8th geometric scale (by height and diameter), high-temperature test facility. A geometric scaling analysis for the VHTR, a time scale analysis of the air-ingress phenomenon, a transient depressurization analysis of the reactor vessel, a hydraulic similarity analysis of the test facility, a heat transfer characterization of the hot plenum, a power scaling analysis for the reactor system, and a design analysis of the containment vessel are discussed.

Development of Tritium Technologies at KAERI

Tritium is formed by neutrons captured from deuterium. If left to accumulate, tritium oxide will become a hazard to the operating staff and public. The primary purpose of a Tritium Removal Facility (TRF) is to reduce tritium concentration in a heavy water moderator. In Korea, operation of a TRF commenced at the Wolsong Nuclear Power Site on July 26th, 2007. Nowadays, KAERI is developing a Very High Temperature Gas Cooled Reactor (VHTR). We have developed a tritium behavior analysis code for the VHTR. We also developed analytical methods for the measurement of food stuffs. Korea shared in the construction of the ITER fuel cycle plant with the EU, Japan, and the US, and is responsible for the supply of an SDS (Tritium Storage and Delivery System). We present the recent progress in the development of tritium storage technology, and safety features of the related system. KAERI has been developing tritium technologies related to the Wolsong TRF, HANARO, VHTR, and nuclear fusion fuel systems. We thus present details on the recent development progress of these tritium systems.