Chang H. Oh

Design Option of Heat Exchanger for the Next Generation Nuclear Plant

The next generation nuclear plant (NGNP), a very high temperature gas-cooled reactor (VHTR) concept, will provide the first demonstration of a closed-loop Brayton cycle at a commercial scale, producing a few hundred megawatts of power in the form of electricity and hydrogen. The power conversion unit for the NGNP will take advantage of the significantly higher reactor outlet temperatures of the VHTRs to provide higher efficiencies than can be achieved with the current generation of light water reactors. Besides demonstrating a system design that can be used directly for subsequent commercial deployment, the NGNP will demonstrate key technology elements that can be used in subsequent advanced power conversion systems for other Generation IV reactors. In anticipation of the design, development, and procurement of an advanced power conversion system for the NGNP, the system integration of the NGNP and hydrogen plant was initiated to identify the important design and technology options that must be considered in evaluating the performance of the proposed NGNP. As part of the system integration of the VHTRs and the hydrogen production plant, the intermediate heat exchanger is used to transfer the process heat from VHTRs to the hydrogen plant. Therefore, the design and configuration of the intermediate heat exchanger are very important. This paper describes analyses of one stage versus two-stage heat exchanger design configurations and simple stress analyses of a printed circuit heat exchanger (PCHE), helical-coil heat exchanger, and shell-and-tube heat exchanger.

Evaluation of working fluids in an indirect combined cycle in a very high temperature gas-cooled reactor

The U.S. Department of Energy and Idaho National Laboratory are developing a very high temperature reactor to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is twofold: (a) efficient, low-cost energy generation and (b) hydrogen production. Although a next-generation plant could be developed as a single-purpose facility, early designs are expected to be dual purpose, as assumed here. A dual-purpose design with a combined cycle of a Brayton top cycle and a bottom Rankine cycle was investigated. An intermediate heat transport loop for transporting heat to a hydrogen production plant was used. Helium, CO2, and a helium-nitrogen mixture were studied to determine the best working fluid in terms of the cycle efficiency. The relative component sizes were estimated for the different working fluids to provide an indication of the relative capital costs. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the cycle were performed to determine the effects of varying conditions in the cycle. This gives some insight into the sensitivity of the cycle to various operating conditions as well as trade-offs between efficiency and component size. Parametric studies were carried out on reactor outlet temperature, mass flow, pressure, and turbine cooling.

RCCS Experiments and Validation for High-Temperature Gas-Cooled Reactor

Abstract An air-cooled helical coil reactor cavity cooling system (RCCS) unit immersed in the water pool was proposed to overcome the disadvantages of the weak cooling ability of an air-cooled RCCS and the complex structure of a water-cooled RCCS for the high-temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool-type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool, and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls.

DEVELOPMENT OF GAMMA CODE AS AN INTEGRATED CODE ANALYZING A COUPLED VERY HIGH TEMPERATURE GAS-COOLED REACTOR AND HYDROGEN PRODUCTION PLANT

Abstract The very high temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 1173 K (900°C) and operational fuel temperatures above 1523 K (1250°C). The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the high-temperature gas-cooled reactor concepts have sufficiently high temperatures to support process heat applications, such as hydrogen production, tar sands, oil shale, desalination, or cogenerative processes, the VHTR’s higher temperatures can be detrimental to safety if a loss-of-coolant accident occurs and causes the mechanical strength degradation of the supporting graphite in the lower plenum. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion or density-gradient–driven stratified flow phenomena and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gases (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Therefore, there was a need to develop a computer code that can be used for VHTR air ingress-related graphite oxidation analyses. Prior to the start of the Republic of Korea/United States International Nuclear Energy Research Initiative collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate an air ingress phenomenon in the VHTR. Therefore, we have worked for the past 3 yr on developing and validating advanced computational methods for simulating air ingress in the VHTR. The Idaho National Laboratory is developing a system integration model of VHTR and hydrogen production plant. GAMMA code is being considered to be an integrated computer tool to analyze the thermal hydraulics of the coupled plant. Computer models for a high-temperature steam electrolysis (HTSE) process were developed and were implemented in an overall system process optimization code, HYSYS. The HTSE model will be implemented into GAMMA code as the integrated computer tool. This paper describes the governing equations and numerical methods used in GAMMA code and presents a portion of verification of the GAMMA code along with turbomachinery models and HTSE models that will be linked to GAMMA code.

Validations of CFD Code for Density-Gradient Driven Air Ingress Stratified Flow

Air ingress into a very high temperature gas-cooled reactor (VHTR) is an important phenomenon to consider because the air oxidizes the reactor core and lower plenum where the graphite structure supports the core region in the gas turbine modular helium reactor (GT-MHR) design, thus jeopardizing the reactor’s safety. Validating the computational fluid dynamics (CFD) code used to analyze the air ingress phenomena is therefore an essential part of the safety analysis and the ultimate computation required for licensing. An experimental data set collected by ETH Zurich on a lock exchange experiment (Grobelbauser et al., Lowe et al. 2002; Lowe et al. 2005; and Shin et al. 2004) was selected for the validation. The experiment was based on a series of lock exchange flows with gases of different density ratios varying from 0.046 to 0.9 in a closed channel of a square cross-section. The focus was on the quantitative measurement of front velocities of the gravity current flows. The experiment results cover the full range of gas intrusions—heavy as well as light—for the gravity current flows in the lock exchange situations. FLUENT CFD code (ANSYS Fluent 2008) was used. The calculated results showed very good agreement with the experimental data. A number of tables and comparison plots are included to summarize the estimated current speeds. The current speed obtained by experimental data was 1.25 m/s and that of the simulation was 1.19 m/s. This result indicates that the deviation of the simulation is only 4.8% that of the experimental data.Copyright

Scoping Analyses on Tritium Permeation to VHTR Integarted Industrial Application Systems

Tritium permeation is a very important current issue in the very high temperature reactor (VHTR) because tritium is easily permeated through high temperature metallic surfaces. Tritium permeations in the VHTR-integrated systems were investigated in this study using the tritium permeation analysis code (TPAC) that was developed by Idaho National Laboratory (INL). The INL TPAC is a numerical tool that is based on the mass balance equations of tritium containing species and hydrogen (i.e. HT, H2, HTO, HTSO4, TI) coupled with a variety of tritium sources, sink, and permeation models. In the TPAC, ternary fission and thermal neutron caption reactions with 6Li, 7Li 10B, 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of tritium and H2 through pipes, vessels, and heat exchangers were considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems including high temperature electrolysis (HTSE) and sulfur-iodine processes.

Air Ingress Analysis: Computational Fluid Dynamics Models

Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy (DOE), is performing research and development that focuses on key phenomena important during potential scenarios that may occur in very high temperature reactors (VHTRs). Phenomena identification and ranking studies to date have ranked an air ingress event, following on the heels of a VHTR depressurization, as important with regard to core safety. Consequently, the development of advanced air-ingress-related models and verification and validation data are a very high priority. Following a loss of coolant and system depressurization incident, air will enter the core of the High Temperature Gas Cooled Reactor through the break, possibly causing oxidation of the core and reflector graphite structure. Simple core and plant models indicate that, under certain circumstances, the oxidation may proceed at an elevated rate with additional heat generated from the oxidation reaction itself. Under postulated conditions of fluid flow and temperature, excessive degradation of lower plenum graphite can lead to a loss of structural support. Excessive oxidation of core graphite can also lead to a release of fission products into the confinement, which could be detrimental to reactor safety. Computational fluid dynamics models developed in this study will improve our understanding of this phenomenon. This paper presents two-dimensional (2-D) and three-dimensional (3-D) computational fluid dynamic (CFD) results for the quantitative assessment of the air ingress phenomena. A portion of the results from density-driven stratified flow in the inlet pipe will be compared with the experimental results.Copyright

Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

ISOTHERMAL AIR-INGRESS VALIDATION EXPERIMENTS

Abstract Idaho National Laboratory has conducted air-ingress experiments as part of a campaign to validate computational fluid dynamics (CFD) calculations for very high-temperature gas-cooled reactor (VHTR) analysis. An isothermal test loop was designed to recreate exchange or stratified flow that occurs in the lower plenum of VHTR after a break in the primary loop allows helium to leak out and reactor building air to enter the reactor core. The experiment was designed to measure stratified flow in the inlet pipe connecting to the lower plenum of the General Atomics gas turbine-modular helium reactor (GT-MHR). Instead of helium and air, brine and sucrose were used as heavy fluids, and water was used as the lighter fluid to create, using scaling laws, the appropriate flow characteristics of the lower plenum immediately after depressurization. These results clearly indicate that stratified flow is established even for very small density differences. Corresponding CFD results were validated with the experimental data. A grid sensitivity study on CFD models was also performed using the Richardson extrapolation and the grid convergence index method for the numerical accuracy of CFD calculations. The calculated current speed showed very good agreement with the experimental data, indicating that current CFD methods are suitable for simulating density gradient stratified flow phenomena in an air-ingress accident.

ISOTHERMAL AIR INGRESS VALIDATION EXPERIMENTS AT IDAHO NATIONAL LABORATORY: DESCRIPTION AND SUMMARY OF DATA

Idaho National Laboratory performed air ingress experiments as part of validating computational fluid dynamics code (CFD). An isothermal stratified flow experiment was designed and set to understand stratified flow phenomena in the very high temperature gas cooled reactor (VHTR) and to provide experimental data for validating computer codes. The isothermal experiment focused on three flow characteristics unique in the VHTR air-ingress accident: stratified flow in the horizontal pipe, stratified flow expansion at the pipe and vessel junction, and stratified flow around supporting structures. Brine and sucrose were used as heavy fluids and water was used as light fluids. The density ratios were changed between 0.87 and 0.98. This experiment clearly showed that a stratified flow between heavy and light fluids is generated even for very small density differences. The code was validated by conducting blind CFD simulations and comparing the results to the experimental data. A grid sensitivity study was also performed based on the Richardson extrapolation and the grid convergence index method for modeling confidence. As a result, the calculated current speed showed very good agreement with the experimental data, indicating that the current CFD methods are suitable for predicting density gradient stratified flow phenomena in the air-ingress accident.