Development of HyPEP, A Hydrogen Production Plant Efficiency Calculation Program

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.

The Very High Temperature Gas-Cooled Reactor (VHTR) coupled to the High Temperature Steam Electrolysis (HTSE) process is one of two reference integrated systems being investigated by the U.S. Department of Energy and Idaho National Laboratory for the production of hydrogen. In this concept the VHTR outlet temperature of 900 °C provides thermal energy and high efficiency electricity for the electrolysis of steam in the HTSE process. In the second reference system the Sulfur Iodine (SI) process is coupled to the VHTR to produce hydrogen thermochemically. In the HyPEP project we are investigating and characterizing these two reference systems with respect to production, operability, and safety performance criteria. Under production, plant configuration and working fluids are being studied for their effect on efficiency. Under operability, control strategies are being developed with the goal of maintaining equipment within operating limits while meeting changes in demand. Safety studies are to investigate plant response for equipment failures. Specific objectives in FY07 were (1) to develop HyPEP Beta and verification and validation (V&V) plan, (2) to perform steady state system integration, (3) to perform parametric studies with various working fluids and power conversion unit (PCU) configurations, (4) the study of design options such as pressure, temperature, etc. (5) to develop a control strategy and (6) to perform transient analyses for plant upsets, control strategy, etc for hydrogen plant with PCU. This report describes the progress made in FY07 in each of the above areas. (1) The HyPEP code numeric scheme and Graphic User Interface have been tested and refined since the release of the alpha version a year ago. (2) The optimal size and design condition for the intermediate heat exchanger, one of the most important components for integration of the VHTR and HTSE plants, was estimated. (3) Efficiency calculations were performed for a variety of working fluids for this reference design. (4) Efficiency improvements over the reference VHTR/HTSE plant were investigated for an alternative design that directly couples a High Temperature Steam Rankin Cycle (HTRC) to the HTSE process. Integration of the VHTR with SI process plants was begun. (5) Plant control studies showed that inventory control in the VHTR plant and flow control in the HTSE plant is effective in maintaining hot-side temperatures near constant during load change. (6) Dynamic calculations showed that thermal transients arising in the chemical plant are strongly damped at the reactor resulting in a stable combined plant.

The tritium concentration in the hydrogen product in Japan's future very high temperature gas-cooled reactor (VHTR) system coupled with a thermochemical water-splitting iodine-sulfur (IS) process (VHTRIS system), named GTHTR300C, was estimated by numerical analysis. The tritium concentration in the hydrogen product significantly depended on undetermined parameters, i.e., the permeabilities of a SO3 decomposer and a H2SO4vaporizer made of SiC. Thus, the estimated tritium concentration in the hydrogen product for the conservative analytical condition ranged from 3.4 × 10−3 Bq/cm3 at STP (38 Bq/g-H2) to 0.18 Bq/cm3 at STP (2,000 Bq/g-H2). By considering the tritium retained by core graphite and the reduction in permeation rate by an oxide film on the heat transfer tube of the IHX and the HI decomposer, the tritium concentration in the hydrogen product decreased to the range from 3.3 × 10−5 Bq/cm3 at STP (0.36 Bq/g-H2) to 5.6 × 10−3 Bq/cm3 at STP (63 Bq/g-H2), which were smaller than those for the conservative analytical condition by factors of about 3.2 × 10−2 and 9.6 × 10−3, respectively. The effectof the helium flow rate in the helium purification system on the tritium concentration in the hydrogen product was also evaluated.

Thermo-economic analysis and optimization of the very high temperature gas-cooled reactor-based nuclear hydrogen production system using copper-chlorine cycle

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.

Experimental studies on the thermal and hydraulic performance of the fuel stack of the VHTR: Part II. HENDEL multi-channel test rig with twelve fuel rods

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

This work is concerning a methodology of Ni-base superalloy development for a very high temperature gas-cooled reactor(VHTR) using design of experiments(DOE) and thermodynamic calculations. Total 32 sets of the Ni-base superalloys with various chemical compositions were formulated based on a fractional factorial design of DOE, and the thermodynamic stability of topologically close-packed(TCP) phases of those alloys was calculated by using the THERMO-CALC software. From the statistical evaluation of the effect of the chemical composition on the formation of TCP phase up to a temperature of 950 oC, which should be suppressed for prolonged service life when it used as the structural components of VHTR, 16 sets were selected for further calculation of the mechanical properties. Considering the yield and ultimate tensile strengths of the selected alloys estimated by using the JMATPRO software, the optimized chemical composition of the alloys for VHTR application, especially intermediate heat exchanger, was proposed for a succeeding experimental study.

Air-ingress events caused by large pipe breaks are important accidents considered in the design of Very High Temperature Gas-Cooled Reactors (VHTRs). A main safety concern for this type of event is the possibility of core collapse following the failure of the graphite support column, which can be oxidized by ingressed air. In this study, the main target is to predict the strength of the oxidized graphite support column. Through compression tests for fresh and oxidized graphite columns, the compressive strength of IG-110 was obtained. The buckling strength of the IG-110 column is expressed using the following empirical straight-line formula: σ cr,buckling = 91.34—1.01 (L/r). Graphite oxidation in Zone 1 is volume reaction and that in Zone 3 is surface reaction. We notice that the ultimate strength of the graphite column oxidized in Zones 1 and 3 only depends on the slenderness ratio and bulk density. Its strength degradation oxidized in Zone 1 is expressed in the following nondimensional form: σ/σ0 = exp(—kd), k = 0.114. We found that the strength degradation of a graphite column, oxidized in Zone 3, follows the above buckling empirical formula as the slenderness of the column changes.

The Very High Temperature gas cooled reactor (VHTR) is one of the GEN IV reactor concepts that have been proposed for thermochemical hydrogen production and other process-heat applications like coal gasification. The USDOE has selected the VHTR for further research and development, aiming to demonstrate emissions-free electricity and hydrogen production at a future time. One of the major safety advantages of the VHTR is the potential for passive decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-side of the RCCS is very similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that has been proposed for the PRISM reactor design. The design and safety analysis of the RVACS have been based on extensive analytical and experimental work performed at ANL. The Natural Convective Shutdown Heat Removal Test Facility (NSTF) at ANL that simulates at full scale the air-side of the RVACS was built to provide experimental support for the design and analysis of the PRISM RVACS system. The objective of this work is to demonstrate that the NSTF facility can be used to generate RCCS experimental data: to validate CFD and systems codes for the analysis of the RCCS; and to support the design and safety analysis of the RCCS.

<p indent=0mm>After the Fukushima nuclear accident in Japan, the safety requirements of nuclear energy systems are constantly increasing. A series of new nuclear energy systems are under development, such as very high-temperature gas-cooled reactor (VHTR), lead-cooled fast reactor (LFR), gas-cooled fast reactor (GFR) and so on. In order to achieve the goals of higher safety, these nuclear systems have put forward higher requirements for nuclear materials. For example, materials working in VHTR and GFR are supposed to withstand higher temperatures, and LFR requires materials to be more corrosion-resistant. Nuclear materials have to meet these requirements under normal operating conditions, as well as in incidental and accidental conditions. Therefore, it is necessary to optimize the existing material system and develop new high-performance materials to meet the future development needs of advanced nuclear power systems. Compared with traditional nuclear materials, carbide ceramics have better mechanical properties under elevated temperatures, stronger irradiation resistance and better comprehensive thermophysical properties. Up to now, great breakthroughs have been made in the research of carbide ceramic materials and some key materials are going to be used in the new-generation nuclear systems. For example, uranium carbide (UC) is gradually replacing traditional uranium dioxide (UO₂), and silicon carbide (SiC) is acting as the main fission product barrier layer of coated fuel particles in HTGR. Other new carbide materials such as zirconium carbide (ZrC) and MAX phase carbide materials are also becoming a new research hotspot in the field of nuclear materials. Compared with other ceramic materials, the fabrication especially its densification of carbide is generally more difficult. In addition, high purity and good neutron properties are also the basic requirements of carbide ceramic materials for nuclear energy. The vapor deposition process and powder metallurgy process have been used to fabricate carbide ceramics for different components in nuclear reactors. Some new preparation methods have been developed to fabricate fuels and structural materials. For example, gelation progress is applied to prepare UCO microsphere fuel, and nano infiltration and transient eutectic (NITE) process is used in the preparation of SiC_f/SiC composites, which is potential for nuclear claddings. The irradiation-induced degradation of nuclear ceramic materials in the reactor is unavoidable. In order to evaluate the performance changes of materials under irradiation conditions, different irradiation methods such as electron irradiation, ion irradiation, proton irradiation and neutron irradiation are used. Due to its unique microstructure and properties, carbide ceramics exhibit different properties from other nuclear materials under irradiation. The performance changes of main nuclear carbide materials after irradiation are summarized, including microstructure evolution, macroscopic irradiation-induced swelling, changes in mechanical and thermal properties, etc. In general, based on the current research progress, the article reviews the application of carbide ceramics in new nuclear power systems and elaborates their service scenarios. Basic properties, preparation methods and irradiation properties of carbide ceramic materials are introduced in detail. The future development directions and the processing techniques are proposed. In the future, carbide ceramic materials should be further developed, and carbide ceramic materials will find diverse applications in the new generation nuclear systems.

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.

The immediate heat exchanger is important for the heat utilization and hydrogen production in a very high-temperature gas-cooled reactor (VHTR). There is a 90-degree elbow passageway in the inlet passageway of immediate heat exchanger, contributing uneven flow distribution in circumference direction in the heat transfer area, which is harmful to the heat transfer tubes aging process. In this study, three inlet passageway optimization designs are put forward and their optimization effect is analyzed according to specific criterion by CFD. As the existence of spiral tube area would have coupling effect on inlet passageway flow, in the computation, the heat transfer area is assumed to be porous medium, in which the effect of heat transfer is taken into account as well. The result shows that installing flow distributer in the inlet passageway could make helium flow on circumference direction most homogenous.

A 2 1/4 Cr-1 Mo steel is a promising candidate material for structural components of the pressure vessel of the experimental very high temperature gascooled reactor (VHTR) in Japan. Since the service temperature of such components is expected to be about 400/sup 0/C, the behavior of the temper and neutron irradiation embrittlements in these chromium-molybdenum steels should be confirmed from the viewpoint of structural integrity. The experimental verification on the degree of the embrittlement due to thermal aging, including the effect of applied stress and neutron irradiation, is described. Steel containing substantial amounts (about 100 ppm) of phosphorus atoms, which are believed to cause the temper embrittlement, showed that applied stress enhanced the embrittlement due to thermal aging. Embrittlement caused by neutron irradiation appears to be minimal in the case of the material containing <1000 ppm of copper as impurity with neutrons irradiated at about 400/sup 0/C.

Leak Before Break (LBB) analysis is widely used to demonstrate system safety of pressurized equipments in nuclear plants. Developing a precise leak rate model for helium flow in complex geometry cracks is a key point for LBB analysis in Very High Temperature gas-cooled Reactor (VHTR). A literature study was firstly carried out concentrating on leak rate models proposed or used in LBB analysis. Helium leak rate experiments named DEDIFAR have been performed using elliptically converging and diverging artificial cracks in order to study the effect of the crack length ratio of a real through wall crack. Computational Fluid Dynamics (CFD) calculations of DEDIFAR experiments are presented to analyze the distributions and tendencies of flow parameters in converging cracks. Then, a parametric study was completed to investigate crack geometry effect on mass flow rate. Based on the results of these studies, the effects of upstream pressure and convergence ratio (inside over outside crack lengths) on mass flow rate are analyzed. The results show that the pressure loss mainly concentrates around the smallest exit crack area, and that when the convergence ratio equals or exceeds 20, flow parameters inside the crack domain remain almost the same.