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STUDY ON AIR INGRESS MITIGATION METHODS IN THE VERY HIGH TEMPERATURE GAS COOLED REACTOR (VHTR)

Description: An air-ingress accident followed by a pipe break is considered as a critical event for a very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break leading to oxidation of the in-core graphite structure. Thus, without mitigation features, this accident might lead to severe exothermic chemical reactions of graphite and oxygen. Under extreme circumstances, a loss of core structural integrity may occur along with excessive release of radiological inventory. Idaho National Laboratory under the auspices of the U.S. Department of Energy is performing research and development (R&D) that focuses on key phenomena important during challenging scenarios that may occur in the VHTR. Phenomena Identification and Ranking Table (PIRT) studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Oh et al. 2006, Schultz et al. 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) requirements are part of the experimental validation plan. This paper discusses about various air-ingress mitigation concepts applicable for the VHTRs. The study begins with identifying important factors (or phenomena) associated with the air-ingress accident by using a root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air-ingress mitigation ideas can be conceptually derived. The main concepts include (1) preventing structural degradation of graphite supporters; (2) preventing local stress concentration in the supporter; (3) preventing graphite oxidation; (4) preventing air ingress; (5) preventing density gradient driven flow; (4) preventing fluid density gradient; (5) preventing fluid temperature gradient; (6) preventing high temperature. Based on the basic concepts listed above, various air-ingress mitigation methods are proposed in this study. Among them, the following two mitigation ideas are extensively investigated using computational ...
Date: March 1, 2011
Creator: Oh, Chang H.
Partner: UNT Libraries Government Documents Department

Numerical Calculations of Wind Flow in a Full-Scale Wind Test Facility

Description: Numerical studies on wind flow around the Texas Tech University (TTU) Wind Engineering Research Field Laboratory (WERFL) building were conducted. The main focus of this paper is wind loads on the TTU building in the INEEL proposed Windstorm Simulation Center. The results are presented in the form of distributions of static pressure, dynamic pressure, pressure coefficients, and velocity vectors on the surface and the vicinity of the TTU building.
Date: June 1, 1999
Creator: Oh, Chang H & Lacey, Jerry Mark
Partner: UNT Libraries Government Documents Department

Development of a Supercritical Carbon Dioxide Brayton Cycle: Improving VHTR Efficiency and Testing Material Compatibility - Final Report

Description: Generation IV reactors will need to be intrinsically safe, having a proliferation-resistant fuel cycle and several advantages relative to existing light water reactor (LWR). They, however, must still overcome certain technical issues and the cost barrier before it can be built in the U.S. The establishment of a nuclear power cost goal of 3.3 cents/kWh is desirable in order to compete with fossil combined-cycle, gas turbine power generation. This goal requires approximately a 30 percent reduction in power cost for stateof-the-art nuclear plants. It has been demonstrated that this large cost differential can be overcome only by technology improvements that lead to a combination of better efficiency and more compatible reactor materials. The objectives of this research are (1) to develop a supercritical carbon dioxide Brayton cycle in the secondary power conversion side that can be applied to the Very-High-Temperature Gas-Cooled Reactor (VHTR), (2) to improve the plant net efficiency by using the carbon dioxide Brayton cycle, and (3) to test material compatibility at high temperatures and pressures. The reduced volumetric flow rate of carbon dioxide due to higher density compared to helium will reduce compression work, which eventually increase plant net efficiency.
Date: June 1, 2006
Creator: Oh, Chang H.
Partner: UNT Libraries Government Documents Department

Scoping Analyses on Tritium Permeation to VHTR Integarted Industrial Application Systems

Description: 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.
Date: March 1, 2011
Creator: Oh, Chang H. & Kim, Eung S.
Partner: UNT Libraries Government Documents Department

ISOTHERMAL AIR INGRESS VALIDATION EXPERIMENTS

Description: Idaho National Laboratory carried out air ingress experiments as part of validating computational fluid dynamics (CFD) calculations. An isothermal test loop was designed and set to understand the stratified-flow phenomenon, which is important as the initial air flow into the lower plenum of the very high temperature gas cooled reactor (VHTR) when a large break loss-of-coolant accident occurs. The unique flow characteristics were focused on the VHTR air-ingress accident, in particular, the flow visualization of the stratified flow in the inlet pipe to the vessel lower plenum of the General Atomic’s Gas Turbine-Modular Helium Reactor (GT-MHR). Brine and sucrose were used as heavy fluids, and water was used to represent a light fluid, which mimics a counter current flow due to the density difference between the stimulant fluids. The density ratios were changed between 0.87 and 0.98. This experiment clearly showed that a stratified flow between simulant fluids was established even for very small density differences. The CFD calculations were compared with 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 . 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.
Date: September 1, 2011
Creator: Oh, Chang H & Kim, Eung S
Partner: UNT Libraries Government Documents Department

TRITIUM PERMEATION AND TRANSPORT IN THE GASOLINE PRODUCTION SYSTEM COUPLED WITH HIGH TEMPERATURE GAS-COOLED REACTORS (HTGRS)

Description: This paper describes scoping analyses on tritium behaviors in the HTGR-integrated gasoline production system, which is based on a methanol-to-gasoline (MTG) plant. In this system, the HTGR transfers heat and electricity to the MTG system. This system was analyzed using the TPAC code, which was recently developed by Idaho National Laboratory. The global sensitivity analyses were performed to understand and characterize tritium behaviors in the coupled HTGR/MTG system. This Monte Carlo based random sampling method was used to evaluate maximum 17,408 numbers of samples with different input values. According to the analyses, the average tritium concentration in the product gasoline is about 3.05×10-3 Bq/cm3, and 62 % cases are within the tritium effluent limit (= 3.7x10-3 Bq/cm3[STP]). About 0.19% of released tritium is finally transported from the core to the gasoline product through permeations. This study also identified that the following four parameters are important concerning tritium behaviors in the HTGR/MTG system: (1) tritium source, (2) wall thickness of process heat exchanger, (3) operating temperature, and (4) tritium permeation coefficient of process heat exchanger. These four parameters contribute about 95 % of the total output uncertainties. This study strongly recommends focusing our future research on these four parameters to improve modeling accuracy and to mitigate tritium permeation into the gasol ine product. If the permeation barrier is included in the future study, the tritium concentration will be significantly reduced.
Date: May 1, 2011
Creator: Oh, Chang H.; Kim, Eung S. & Patterson, Mike
Partner: UNT Libraries Government Documents Department

Air Ingress Analyses on a High Temperature Gas-Cooled Reactor

Description: A primary-pipe break accident is one of the design-basis accidents of a high-temperature gas-cooled reactor (HTGR). When this accident occurs, air is anticipated to enter the reactor core from the break and oxidize the in-core graphite structure in the modular pebble bed reactor (MPBR). This paper presents the results of the graphite oxidation model developed as part of the Idaho National Engineering and Environmental Laboratory's Direct Research and Development effort. Although gas reactors have been tried in the past with limited success, the innovations of modularity and integrated state-ofart control systems coupled with improved fuel design and a pebble bed core make this design potentially very attractive from an economic and technical perspective. A schematic diagram on a reference design of the MPBR has been established on a major component level (INEEL & MIT, 1999). Steady-state and transient thermal hydraulics models will be produced with key parameters established for these conditions at all major components. Development of an integrated plant model to allow for transient analysis on a more sophisticated level is now being developed. In this paper, preliminary results of the hypothetical air ingress are presented. A graphite oxidation model was developed to determine temperature and the control mechanism in the spherical graphite geometry.
Date: November 1, 2001
Creator: Oh, Chang H; Moore, Richard Leroy; Merrill, Brad Johnson & Petti, David Andrew
Partner: UNT Libraries Government Documents Department

Development of Tritium Permeation Analysis Code (TPAC)

Description: Idaho National Laboratory developed the Tritium Permeation Analysis Code (TPAC) for tritium permeation in the Very High Temperature Gas Cooled Reactor (VHTR). All the component models in the VHTR were developed and were embedded into the MATHLAB SIMULINK package with a Graphic User Interface. The governing equations of the nuclear ternary reaction and thermal neutron capture reactions from impurities in helium and graphite core, reflector, and control rods were implemented. The TPAC code was verified using analytical solutions for the tritium birth rate from the ternary fission, the birth rate from 3He, and the birth rate from 10B. This paper also provides comparisons of the TPAC with the existing other codes. A VHTR reference design was selected for tritium permeation study from the reference design to the nuclear-assisted hydrogen production plant and some sensitivity study results are presented based on the HTGR outlet temperature of 750 degrees C.
Date: October 1, 2010
Creator: Kim, Eung S.; Oh, Chang H. & Patterson, Mike
Partner: UNT Libraries Government Documents Department

Method and System for the Production of Hydrogen at Reduced VHTR Outlet Temperatures

Description: The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility dedicated to hydrogen production, early designs are expected to be dual purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor with electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. The integrated system of a Very High Temperature Reactor (VHTR) and a High Temperature Steam Electrolysis (HTSE) hydrogen production plant is being investigated and this system, as it is currently envisioned, will produce hydrogen by utilizing a highly efficient VHTR with a VHTR outlet temperature of 900°C to supply the necessary energy and electricity to the HTSE unit. Though the combined system may produce hydrogen and electricity with high efficiency, the choices of materials that are suitable for use at 900°C are limited due to high-temperature strength, corrosion, and durability (creep) considerations. The lack of materials that are ASME (American Society of Mechanical Engineers) code-certified at these temperatures is also a problem, and is a barrier to commercial deployment. If the current system concept can be modified to produce hydrogen with comparable efficiency at lower temperatures, then the technical barriers related to materials selection and use might be eliminated, and the integrated system may have a much greater probability of succeeding at ...
Date: October 1, 2009
Creator: Oh, Chang H. & Kim, Eung S.
Partner: UNT Libraries Government Documents Department

Development and Verification of Tritium Analyses Code for a Very High Temperature Reactor

Description: A tritium permeation analyses code (TPAC) has been developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in the VHTR systems including integrated hydrogen production systems. A MATLAB SIMULINK software package was used for development of the code. The TPAC is based on the mass balance equations of tritium-containing species and a various form of hydrogen (i.e., HT, H2, HTO, HTSO4, and TI) coupled with a variety of tritium source, sink, and permeation models. In the TPAC, ternary fission and neutron 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 HT and H2 through pipes, vessels, and heat exchangers were importantly considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems including both high-temperature electrolysis and sulfur-iodine process. The TPAC has unlimited flexibility for the system configurations, and provides easy drag-and-drops for making models by adopting a graphical user interface. Verification of the code has been performed by comparisons with the analytical solutions and the experimental data based on the Peach Bottom reactor design. The preliminary results calculated with a former tritium analyses code, THYTAN which was developed in Japan and adopted by Japan Atomic Energy Agency were also compared with the TPAC solutions. This report contains descriptions of the basic tritium pathways, theory, simple user guide, verifications, sensitivity studies, sample cases, and code tutorials. Tritium behaviors in a very high temperature reactor/high temperature steam electrolysis system have been analyzed by the TPAC based on the reference indirect parallel configuration proposed by Oh et al. (2007). This analysis showed that only 0.4% of tritium released from the core is transferred to the product hydrogen. The amount of tritium in the product ...
Date: September 1, 2009
Creator: Oh, Chang H. & Kim, Eung S.
Partner: UNT Libraries Government Documents Department

Heat Exchanger Design Options and Tritium Transport Study for the VHTR System

Description: This report presents the results of a study conducted to consider heat exchanger options and tritium transport in a very high temperature reactor (VHTR) system for the Next Generation Nuclear Plant Project. The heat exchanger options include types, arrangements, channel patterns in printed circuit heat exchangers (PCHE), coolant flow direction, and pipe configuration in shell-and-tube designs. Study considerations include: three types of heat exchanger designs (PCHE, shell-and-tube, and helical coil); single- and two-stage unit arrangements; counter-current and cross flow configurations; and straight pipes and U-tube designs in shell-and-tube type heat exchangers. Thermal designs and simple stress analyses were performed to estimate the heat exchanger options, and the Finite Element Method was applied for more detailed calculations, especially for PCHE designs. Results of the options study show that the PCHE design has the smallest volume and heat transfer area, resulting in the least tritium permeation and greatest cost savings. It is theoretically the most reliable mechanically, leading to a longer lifetime. The two-stage heat exchanger arrangement appears to be safer and more cost effective. The recommended separation temperature between first and second stages in a serial configuration is 800oC, at which the high temperature unit is about one-half the size of the total heat exchanger core volume. Based on simplified stress analyses, the high temperature unit will need to be replaced two or three times during the plant’s lifetime. Stress analysis results recommend the off-set channel pattern configuration for the PCHE because stress reduction was estimated at up to 50% in this configuration, resulting in a longer lifetime. The tritium transport study resulted in the development of a tritium behavior analysis code using the MATLAB Simulink code. In parallel, the THYTAN code, previously performed by Ohashi and Sherman (2007) on the Peach Bottom data, was revived and verified. The 600 MWt ...
Date: September 1, 2008
Creator: Oh, Chang H. & Kim, Eung S.
Partner: UNT Libraries Government Documents Department

Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant

Description: The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood.
Date: April 1, 2008
Creator: Oh, Chang H.; Kim, Eung Soo & Sherman, Steven
Partner: UNT Libraries Government Documents Department

Development of Tritium Permeation Analysis Code and Tritium Transport in a High Temperature Gas-Cooled Reactor Coupled with Hydrogen Production System

Description: Abstract – A tritium permeation analyses code (TPAC) was developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in very high temperature reactor (VHTR) systems, including integrated hydrogen production systems. A MATLAB SIMULINK software package was used in developing the code. The TPAC is based on the mass balance equations of tritium-containing species and various forms of hydrogen coupled with a variety of tritium sources, sinks, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, and 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 and sulfur-iodine processes.
Date: June 1, 2010
Creator: Oh, Chang H.; Kim, Eung S. & Patterson, Mike
Partner: UNT Libraries Government Documents Department

Isothermal Air Ingress Validation Experiments at Idaho National Laboratory: Description and Summary of Data

Description: 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.
Date: September 1, 2010
Creator: Oh, Chang H. & Kim, Eung S.
Partner: UNT Libraries Government Documents Department

FINAL REPORT on Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

Description: The U.S. Department of Energy is performing research and development that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Generation IV very high temperature reactor (VHTR). Phenomena Identification and Ranking studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important. Consequently, the development of advanced air ingress-related models and verification & validation are of very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air ingress will occur through the break, leading to oxidation of the in-core graphite structure and fuel. This study indicates that depending on the location and the size of the pipe break, the air ingress phenomena are different. In an effort to estimate the proper safety margin, experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model are required. It will also require effective strategies to mitigate the effects of oxidation, eventually. This 3-year project (FY 2008–FY 2010) is focused on various issues related to the VHTR air-ingress accident, including (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the core bottom structures, (d) structural tests of the oxidized core bottom structures, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i) verification and validation of the coupled models.
Date: January 1, 2011
Creator: Oh, Chang H.; Kim, Eung S.; NO, Hee C. & Cho, Nam Z.
Partner: UNT Libraries Government Documents Department

POWER CYCLE AND STRESS ANALYSES FOR HIGH TEMPERATURE GAS-COOLED REACTOR

Description: The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with three turbines and four compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with three stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. 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 three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles ...
Date: May 1, 2007
Creator: Oh, Chang H.; Davis, Cliff; Hawkes, Brian D. & Sherman, Steven R.
Partner: UNT Libraries Government Documents Department

Stress and Heat Transfer Analyses for Different Channel Arrangements of PCHE

Description: Stress and heat transfer analyses are being performed on the different channel arrangements of Printed Circuit Heat Exchanger (PCHE) proposed for application of VHTRs using ABAQUS [ABAQUS, 2007] and COMSOL [COMSOL, 2007], respectively. The work is being done to determine the configuration that would result in minimum stress for the same heat performance. This paper discusses the effects of shifting the coolant channels in every other row to reduce stress.
Date: November 1, 2008
Creator: Lim, Jong B.; Shrake, Robert G.; Kim, Eung S. & Oh, Chang H.
Partner: UNT Libraries Government Documents Department

Computational Fluid Dynamics Analyses on Very High Temperature Reactor Air Ingress

Description: A preliminary computational fluid dynamics (CFD) analysis was performed to understand density-gradient-induced stratified flow in a Very High Temperature Reactor (VHTR) air-ingress accident. Various parameters were taken into consideration, including turbulence model, core temperature, initial air mole-fraction, and flow resistance in the core. The gas turbine modular helium reactor (GT-MHR) 600 MWt was selected as the reference reactor and it was simplified to be 2-D geometry in modeling. The core and the lower plenum were assumed to be porous bodies. Following the preliminary CFD results, the analysis of the air-ingress accident has been performed by two different codes: GAMMA code (system analysis code, Oh et al. 2006) and FLUENT CFD code (Fluent 2007). Eventually, the analysis results showed that the actual onset time of natural convection (~160 sec) would be significantly earlier than the previous predictions (~150 hours) calculated based on the molecular diffusion air-ingress mechanism. This leads to the conclusion that the consequences of this accident will be much more serious than previously expected.
Date: July 1, 2009
Creator: Oh, Chang H; Kim, Eung S.; Schultz, Richard; Petti, David & Kang, Hyung S.
Partner: UNT Libraries Government Documents Department

Effect of Reacting Surface Density on the Overall Graphite Oxidation Rate

Description: Graphite oxidation in an air-ingress accident is presently a very important issue for the reactor safety of the very high temperature gas cooled-reactor (VHTR), the concept of the next generation nuclear plant (NGNP) because of its potential problems such as mechanical degradation of the supporting graphite in the lower plenum of the VHTR might lead to core collapse if the countermeasure is taken carefully. The oxidation process of graphite has known to be affected by various factors, including temperature, pressure, oxygen concentration, types of graphite, graphite shape and size, flow distribution, etc. However, our recent study reveals that the internal pore characteristics play very important roles in the overall graphite oxidation rate. One of the main issues regarding graphite oxidation is the potential core collapse problem that may occur following the degradation of graphite mechanical strength. In analyzing this phenomenon, it is very important to understand the relationship between the degree of oxidization and strength degradation. In addition, the change of oxidation rate by graphite oxidation degree characterization by burn-off (ratio of the oxidized graphite density to the original density) should be quantified because graphite strength degradation is followed by graphite density decrease, which highly affects oxidation rates and patterns. Because the density change is proportional to the internal pore surface area, they should be quantified in advance. In order to understand the above issues, the following experiments were performed: (1)Experiment on the fracture of the oxidized graphite and validation of the previous correlations, (2) Experiment on the change of oxidation rate using graphite density and data collection, (3) Measure the BET surface area of the graphite. The experiments were performed using H451 (Great Lakes Carbon Corporation) and IG-110 (Toyo Tanso Co., Ltd) graphite. The reason for the use of those graphite materials is because their chemical and mechanical characteristics ...
Date: May 1, 2009
Creator: Oh, Chang H.; Kim, Eung; Lim, Jong; Schultz, Richard & Petti, David
Partner: UNT Libraries Government Documents Department

Steam Generator Component Model in a Combined Cycle of Power Conversion Unit for Very High Temperature Gas-Cooled Reactor

Description: The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP), Very High Temperature Gas-Cooled Reactor (VHTR) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. A combined cycle is considered as one of the power conversion units to be coupled to the very high-temperature gas-cooled reactor (VHTR). The combined cycle configuration consists of a Brayton top cycle coupled to a Rankine bottoming cycle by means of a steam generator. A detailed sizing and pressure drop model of a steam generator is not available in the HYSYS processes code. Therefore a four region model was developed for implementation into HYSYS. The focus of this study was the validation of a HYSYS steam generator model of two phase flow correlations. The correlations calculated the size and heat exchange of the steam generator. To assess the model, those calculations were input into a RELAP5 model and its results were compared with HYSYS results. The comparison showed many differences in parameters such as the heat transfer coefficients and revealed the different methods used by the codes. Despite differences in approach, the overall results of heat transfer were in good agreement.
Date: June 1, 2007
Creator: Oh, Chang H; Han, James; Barner, Robert & Sherman, Steven R
Partner: UNT Libraries Government Documents Department

Development of GAMMA Code and Evaluation for a Very High Temperature gas-Cooled Reactor

Description: 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 9000C and operational fuel temperatures above 12500C. 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 (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion 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 gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. This paper will also include what improvements will be made in the Gamma code for the VHTR.
Date: June 1, 2007
Creator: Oh, Chang H; Lim, H.S.; Kim, E.S. & NO, H.C.
Partner: UNT Libraries Government Documents Department

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

Description: Air ingress into a very high temperature gas-cooled reactor (VHTR) is an important phenomena 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 (GTMHR) 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
Date: May 1, 2010
Creator: Oh, Chang H.; Kim, Eung S.; Schultz, Richard & Petti, David
Partner: UNT Libraries Government Documents Department

Computational Assessment of the GT-MHR Graphite Core Support Structural Integrity in Air-Ingress Accident Condition

Description: The objective of this project was to perform stress analysis for graphite support structures of the General Atomics’ 600 MWth GT-MHR prismatic core design using ABAQUS ® (ver. 6.75) to assess their structural integrity in air-ingress accident conditions where the structure weakens over time due to oxidation damages. The graphite support structures of prismatic type GT-MHR was analyzed based on the change of temperature, burn-off and corrosion depth during the accident period predicted by GAMMA, a multi-dimensional gas multi-component mixture analysis code developed in the Republic of Korea (ROK)/United States (US) International –Nuclear Engineering Research Initiative (I-NERI) project. Both the loading and thermal stresses were analyzed, but the thermal stress was not significant, leaving the loading stress to be the major factor. The mechanical strengths are exceeded between 11 to 11.5 days after loss-of-coolant-accident (LOCA), corresponding to 5.5 to 6 days after the start of natural convection.
Date: October 1, 2008
Creator: Lim, Jong B.; Kim, Eung S.; Oh, Chang H.; Schultz, Richard R. & Petti, David A.
Partner: UNT Libraries Government Documents Department

AIR INGRESS ANALYSIS: COMPUTATIONAL FLUID DYNAMIC MODELS

Description: The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, 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 in-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 the lower plenum graphite can lead to a loss of structural support. Excessive oxidation of core graphite can also lead to the release of fission products into the confinement, which could be detrimental to a reactor safety. Computational fluid dynamic model developed in this study will improve our understanding of this phenomenon. This paper presents two-dimensional and three-dimensional CFD results for the quantitative assessment of the air ingress phenomena. A portion of results of the density-driven stratified flow in the inlet pipe will be compared with results of the experimental results.
Date: August 1, 2010
Creator: Oh, Chang H.; Kim, Eung S.; Schultz, Richard; Gougar, Hans; Petti, David & Kang, Hyung S.
Partner: UNT Libraries Government Documents Department