High Temperature Engineering Test Reactor Research Articles

Numerical Study on Tritium Behavior by Using Isotope Exchange Reactions in Thermochemical Water-Splitting Iodine—Sulfur Process

One potential problem in the hydrogen production system coupled with the high-temperature gascooled reactor (HTGR) is transmission of tritium from the primary coolant to the product hydrogen by permeation through the heat transfer tubes. Tritium accumulation in the process chemicals in the components of a hydrogen plant, a thermochemical water-splitting iodine-sulfur (IS) process, will also be a critical issue in seeking to license the hydrogen plant as a non-nuclear plant in the future. A numerical analysis model for tritium behavior in the IS process was developed by considering the isotope exchange reactions between tritium and the hydrogen-containing process chemicals, i.e., H2O, H2SO4 and HI. The tritium activity concentration in the IS process coupled with the high-temperature engineering test reactor (HTTR), the HTTR-IS system, was preliminarily evaluated in regard to the effects of some indeterminate parameters, i.e., equilibrium constants of the isotope exchange reactions, permeability of tritium through heat transfer tubes, tritium and hydrogen concentrations in the secondary helium coolant, and the leak rate from the secondary coolant loop. The results describing how the tritium activity concentration changes with variations in these parameters and which component has the maximum tritium activity concentration in the IS process are described in this paper.

Conceptual design of hydrogen production system with thermochemical water-splitting iodine–sulphur process utilizing heat from the high-temperature gas-cooled reactor HTTR

Japan Atomic Energy Agency (JAEA) has launched the preliminary design of a hydrogen production system based on Japan's first high-temperature gas-cooled reactor high temperature engineering test reactor (HTTR) since 2005. The selected method of hydrogen production is the thermochemical water-splitting iodine–sulphur (IS) process. The conceptual design of the HTTR–IS system is described in this paper. Using the secondary helium loop of the HTTR for coupling the hydrogen production system to the reactor, the possibility of constructing the IS process as a non-nuclear grade or conventional chemical plant is investigated for the available structure of the HTTR–IS system and the proposed heat mass balance. A hydrogen production rate of about 1100 Nm 3 / h and a thermal efficiency of 44% are shown in the process flowsheet evaluation for the HTTR–IS system.

JAEA'S VHTR FOR HYDROGEN AND ELECTRICITY COGENERATION : GTHTR300C

Design study on the Gas Turbine High Temperature Reactor 300-Cogeneration (GTHTR300C) aiming at producing both electricity by a gas turbine and hydrogen by a thermochemical water splitting method (IS process method) has been conducted. It is expected to be one of the most attractive systems to provide hydrogen for fuel cell vehicles after 2030. The GTHTR300C employs a block type Very High Temperature Reactor (VHTR) with thermal power of 600MW and outlet coolant temperature of . The intermediate heat exchanger (IHX) and the gas turbine are arranged in series in the primary circuit. The IHX transfers the heat of 170MW to the secondary system used for hydrogen production. The balance of the reactor thermal power is used for electricity generation. The GTHTR300C is designed based on the existing technologies of the High Temperature Engineering Test Reactor (HTTR) and helium turbine power conversion and on the technologies whose development have been well under way for IS hydrogen production process so as to minimize cost and risk of deployment. This paper describes the original design features focusing on the plant layout and plant cycle of the GTHTR300C together with present development status of the GTHTR300, IHX, etc. Also, the advantage of the GTHTR300C is presented.

ICONE15-10158 CORE DYNAMICS ANALYSIS FOR REACTIVITY INSERTION AND LOSS OF COOLANT FLOW TESTS USING THE HTTR

The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950℃ (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed. The reactor power decreases to decay heat level from the maximum reactor power of 30 MW due to the negative reactivity feedback effect of the core. Although the reactor power becomes critical again about five hours later, the peak power value is merely 2 MW. It was confirmed that by using the developed code, it is possible to not only analyze the reactor core dynamics but also simulate the core dynamics during the abnormal events postulated in the HTGR safety analysis.

ICONE15-10222 IMPROVEMENT OF ANALYSIS TECHNOLOGIES FOR HTGR BY USING THE HTTR DATA

The Very High Temperature Reactor (VHTR) system, which is one of generation IV reactors, is the high temperature gas-cooled reactor (HTGR) with capabilities of hydrogen production and high efficiency electricity generation. The High Temperature Engineering Test Reactor (HTTR) is the first HTGR in Japan. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 950℃ in April, 2004 during the "rise-to-power tests" confirming the reactor performance. The safety demonstration tests by using the HTTR started from 2002 and are under going to demonstrate inherent safety features of HTGRs. The experimental data obtained in these tests are inevitable to design the VHTR with the high cost performance. The analytical models validated through these tests in the HTTR are applicable to precise simulation of an HTGR performance and can contribute to the research and development of the VHTR.

高温ガス炉における領域別温度係数の適用効果および伝熱計算モデルの改良

A HTTR (High Temperature Engineering Test Reactor), which has a thermal output of 30 MW, a coolant inlet temperature of 395°C and a coolant outlet temperature of 850°C/950°C, is a first high-temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accidents. The safety demonstration tests such as the reactivity insertion and the coolant flow reduction tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of HTGRs. These tests demonstrate that the rapid increase or decrease in the reactor power is restrained by only the negative reactivity feedback effect without an operation of the reactor power control system, and the temperature transient of the reactor is slow. A one-point core dynamics approximation with one fuel channel model could not simulate accurately the reactor power behavior. On the other hand, an original new method using region temperature coefficients and a connection between some fuel channel models and whole core component model for calculating radial heat transfer in the core is adopted. It is crucial to evaluate this method precisely to simulate a performance of HTGR with showing excellent inherent safety features under the reactivity incident by gas circulator tripping error.

ＨＴＴＲ‐ＩＳ水素製造システムにおける安全設計の考え方

Japan Atomic Energy Agency (JAEA) has been conducting an R&D work on the VHTR reactor system and IS hydrogen production system to realize hydrogen production using nuclear heat. As a part of this activity, JAEA is planning to connect an IS test system to the High Temperature Engineering Test Reactor (HTTR) to demonstrate its technical feasibility. This paper proposes a fundamental philosophy on the safety design of the HTTR-IS hydrogen production system including the methodology to select postulated abnormal events and its event sequences and to define safety functions of the IS system to ensure the reactor safety. Also the measure to clarify the IS system as non-reactor system is proposed.

ICONE15-10331 Study on Corrosion Resistance of Glass Lining and Gold Plating in High Temperature Sulfuric Acid for Thermochemical Hydrogen Production IS Process

In the framework of the HTTR (High Temperature Engineering Test Reactor) project, Japan Atomic Energy Agency has been conducting R&D on thermo-chemical Iodine-Sulfur (IS) process for large scale hydrogen production. Since the IS process constitutes corrosive environments, screening of the materials of construction is very important. Corrosion resistance of lining materials, were examined in 47, 75, 90wt% sulfuric acid for 5 hours at temperatures ranging from 200℃ to 400℃ under inert gas pressure of 2 MPaG. Gold plating materials showed good corrosion resistance irrespective of the base metal, Alloy B2 and SUS304. Soda glass for glass lining material was corroded in 47wt% sulfuric acid at 200℃. However, it showed good corrosion resistance in 75wt% sulfuric acid at 300℃ and in 90wt% sulfuric acid at 400℃.

0616 高温ガス炉の冷却材流量喪失時の炉容器冷却システムに関する解析研究(S45-4 原子炉システムおよびその要素技術(4),S45 原子炉システムおよびその要素技術)

Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are being performed to verify the inherent safety features and to validate the numerical code for the safety assessment of the VHTR (Very High Temperature Reactor). The test of loss of coolant flow as one of safety demonstration tests is carried out by tripping of the helium gas circulators. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the vessel cooling system (VCS) during the loss of coolant flow. The temperature distribution of the RPV and surrounding concrete structure were obtained using a commercially available analysis code STAR-CD. The effect of thermal radiation from the RPV was evaluated using the analytical and experimental results. It was found that it is important to take account of the effect of thermal radiation in the transient analysis to evaluate the temperature change of the concrete accurately.

高温ガス炉水素電力コジェネレーションシステム用中間熱交換器の設計，（Ｉ）伝熱管のスペック選定および１次応力評価

Japan Atomic Energy Agency (JAEA) has been carrying out design studies of the Gas Turbine High Temperature Reactor 300 for Cogeneration (GTHTR300C). One of the key components in the GTHTR300C is an intermediate heat exchanger (IHX). The IHX for the GTHTR300C is rated 170MWt and designed on the basis of the shell & tube type IHX adopted in the High Temperature engineering Test Reactor (HTTR). As a conceptual design, we selected the heat exchanger tube size which is based on heat transfer calculation, designed conceptual structure, and evaluated the primary stress in design condition. By the results of this study, the technical feasibility of the large size shell & tube type IHX for the GTHTR300C was verified.

ニューラルネットワークを用いた高温工学試験研究炉（ＨＴＴＲ）の運転監視

ニューラルネットワークを用いた高温工学試験研究炉（ＨＴＴＲ）の運転監視

G222 高温ガス炉の圧力容器温度分布に対する熱放射の影響(ふく射)

Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are being performed to verify the inherent safety features and to validate the numerical code for the safety assessment of the VHTR (Very High Temperature Reactor). The test of loss of coolant flow as one of safety demonstration tests is carried out by tripping of the helium gas circulators. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the vessel cooling system (VCS) during the loss of coolant flow. The temperature distribution of the RPV and surrounding concrete structure were obtained using a commercially available analysis code STAR-CD. It is necessary to consider the effect of thermal radiation from the RPV for evaluation of temperature of the VCS and concrete vessel.

ICONE15-10178 DESIGN STUDY OF PILOT TEST PLANT FOR HYDROGEN PRODUCTION BY THERMO-CHEMICAL WATER SPLITTING IS PROCESS

The Japan Atomic Energy Agency (JAEA) has been conducting the HTTR (High-Temperature Engineering Test Reactor) project from the view point of establishing technology base on HTGR and also on the IS process. In the IS process, raw material, water, is to be reacted with iodine (I_2) and sulfur dioxide (SO_2) to produce hydrogen iodide (HI) and sulfuric acid (H_2SO_4), the so-called Bunsen reaction, which are then decomposed endothermically to produce hydrogen (H_2) and oxygen (O_2), respectively. Iodine and sulfur dioxide produced in the decomposition reactions can be used again as the reactants in the Bunsen reaction. In JAEA, continuous hydrogen production was demonstrated with the hydrogen production rate of about 30 NL/hr for one week using a bench-scale test apparatus made of glass. Based on the test results and know-how obtained through the bench-scale tests, a pilot test plant that can produce hydrogen of about 30 Nm^3/hr is being designed. The test plant will be fabricated with industrial materials such as glass coated steel, SiC ceramics etc, and operated under high pressure condition up to 2 MPa. In parallel to the design study, key components of the IS process such as the sulfuric acid (H_2SO_4) and the sulfur trioxide (SO_3) decomposers working under-high temperature corrosive environments have been designed and test-fabricated to confirm their fabricability. Also, other R&D's are under way such as steady state simulation of pilot system, corrosion resistant seal and piping technology. This paper describes present status of these activities.

Numerical Study on Tritium Behavior by Using Isotope Exchange Reactions in Thermochemical Water-Splitting Iodine–Sulfur Process

One potential problem in the hydrogen production system coupled with the high-temperature gascooled reactor (HTGR) is transmission of tritium from the primary coolant to the product hydrogen by permeation through the heat transfer tubes. Tritium accumulation in the process chemicals in the components of a hydrogen plant, a thermochemical water-splitting iodine-sulfur (IS) process, will also be a critical issue in seeking to license the hydrogen plant as a non-nuclear plant in the future. A numerical analysis model for tritium behavior in the IS process was developed by considering the isotope exchange reactions between tritium and the hydrogen-containing process chemicals, i.e., H2O, H2SO4 and HI. The tritium activity concentration in the IS process coupled with the high-temperature engineering test reactor (HTTR), the HTTR-IS system, was preliminarily evaluated in regard to the effects of some indeterminate parameters, i.e., equilibrium constants of the isotope exchange reactions, permeability of tritium through heat transfer tubes, tritium and hydrogen concentrations in the secondary helium coolant, and the leak rate from the secondary coolant loop. The results describing how the tritium activity concentration changes with variations in these parameters and which component has the maximum tritium activity concentration in the IS process are described in this paper.

A design-phase PSA of a nuclear-powered hydrogen plant

A probabilistic safety assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a high-temperature gas-cooled nuclear reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute's (JAERI) High Temperature Engineering Test Reactor (HTTR) prototype in Japan. The objective of this paper is to show how the PSA can be used for improving the design of the coupled plants. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The results of the PSA show that the average frequency of an accident at this complex that could affect the population is 7 × 10 −8 year −1 which is divided into the various end states. The dominant sequences are those that result in a methane explosion and occur with a frequency of 6.5 × 10 −8 year −1, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR.

Development of control technology for HTTR hydrogen production system with mock-up test facility: System controllability test for loss of chemical reaction

The Japan Atomic Energy Agency has been planning the demonstration test of hydrogen production with the High Temperature Engineering Test Reactor (HTTR). In a HTTR hydrogen production system (HTTR-H2), it is required to control a primary helium temperature within an allowable value at a reactor inlet to prevent a reactor scram. A cooling system for a secondary helium with a steam generator (SG) and a radiator is installed at the downstream of a chemical rector in a secondary helium loop in order to mitigate the thermal disturbance caused by the hydrogen production system. Prior to HTTR-H2, the simulation test with a mock-up test facility has been carried out to establish the controllability on the helium temperature using the cooling system against the loss of chemical reaction. It was confirmed that the fluctuations of the helium temperature at chemical reactor outlet, more than 200 K, at the loss of chemical reaction could be successfully mitigated within the target of ±10 K at SG outlet. A dynamic simulation code of the cooling system for HTTR-H2 was verified with the obtained test data.

Development of Control Technology for the HTTR Hydrogen Production System

Japan Atomic Energy Research Institute has performed the research and development of hydrogen production using the high-temperature engineering test reactor (HTTR). One of the key issues for the HTTR hydrogen production system is the development of control technology for stable operation. A thermal load absorber concept using a steam generator installed downstream of a reformer is proposed to mitigate a variation of helium temperature. Thermal-hydraulic analyses for the start-up operation and the suspension of the feed gas supply to the reformer are carried out. These results show that a large variation of the reformer outlet helium temperature takes place because of a change of the feed gas flow rate. However, the steam generator can mitigate the variation of the helium temperature. It is clarified that the HTTR can continue normal operation independently of the feed gas flow rate.

HTTRにおける制御棒引抜き試験解析および炉心定数原子炉出力30～60%

The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395°C and coolant outlet temperature of 850°C/950°C, is a first high temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accident. Safety demonstration tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of the HTGR. The reactivity insertion test demonstrates that rapid increase of reactor power during withdrawal of the control rod is restrained by only the negative reactivity feedback effect without operation of the reactor power control system, and the kemperature transient of the reactor is slow. A one-point core dynamics approximation with one fuel channel, one fuel temperature coefficient and one moderator coefficient could not simulate accurately the reactor power behavior. On the other hand, an original new method using some temperature coefficients for regions in the core was very effective. It is crucial to evaluate this method precisely to simulate a performance of HTGR with showing excellent inherent safety features under the reactivity initiated incident by control rod withdrawal error. Moreover, all experimental data of reactivity insertion tests and analytical results are released.

高温ガス炉ガスタービン発電システム(GTHTR300)の経済性評価

Japan Atomic Energy Research Institute (JAERI) has been developing a graphite moderate and helium cooled High Temperature Gas-cooled Reactor (HTGR) with gas turbine, the GTHTR300 based on experience gained in development and operations of the High Temperature Engineering Test Reactor (HTTR) in JAERI. The GTHTR300 is a simplified and economical power plant with a high level of safety characteristics and a high plant efficiency of approximately 46%. Cost evaluation for plant construction and power generation is studied in order to clarify the economical feasibility of the GTHTR300. The construction cost is estimated to be about 200 thousands Yen/kWe. The power generation cost is estimated to be about 3.8Yen/kWh by the conditions of 90% load factor and 3% discount rate. The economical feasibility of the GTHTR300 is certified. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan.

ニューラルネットワークを用いた反応度監視による高温工学試験研究炉（HTTR）の異常検知手法の開発

ニューラルネットワークを用いた反応度監視による高温工学試験研究炉（HTTR）の異常検知手法の開発