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Journal Articles

High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950$$^{circ}$$C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.

Journal Articles

Experimental study on application of heat transfer enhancement method using porous material with high porosity

Takeda, Tetsuaki; Ichimiya, Koichi*; Yamauchi, Daiki*

Nihon Kikai Gakkai Rombunshu, B, 74(748), p.189 - 196, 2008/12

The objectives of this study are to clarify performances of a heat transfer enhancement method using porous material with high porosity and to discuss applicability of this method for a gas heating type steam reformer. A heat transfer experiment has been performed using a simulated steam reformer apparatus to obtain characteristics of heat transfer and pressure drop. From the results obtained in this experiment, this method showed a good performance regarding heat transfer. It was also found that an enhancement of heat transfer using porous material with high porosity is further improved under the high temperature conditions as compared with the other methods. There is the possibility of the adoption of this method to the steam reformer of the hydrogen production system connected to a high temperature gas-cooled reactor.

JAEA Reports

Development of reactor kinetics analysis code

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

JAEA-Data/Code 2008-015, 94 Pages, 2008/10

JAEA-Data-Code-2008-015.pdf:5.14MB

A reactor kinetics analysis code called the TAC/BLOOST code was developed for High Temperature Gas-Cooled Reactors (HTGRs). The TAC/BLOOST code can use a divided core model with region temperature coefficients. In this study, a validation of the TAC/BLOOST code was conducted with the experimental data of the High Temperature Engineering Test Reactor (HTTR). As a result, some improved points of the property changes of the fuels and the structures according to the burnup effect, the gap changes among the blocks, the gap changes between the fuel compact and the graphite sleeve, and the cross section changes of the coolant were clarified. Moreover, prediction analyses of the gas circulator tripping tests can be showed within 3% accuracy.

Journal Articles

Validation of IHX temperature calculation code for future HTGR

Nakagawa, Shigeaki; Saikusa, Akio; Tochio, Daisuke; Takeda, Tetsuaki*

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

The intermediate heat exchanger (IHX) is one of key components in the very high temperature reactor (VHTR) system. The IHX is a helium-helium heat exchanger and the secondary hot helium gas heated up to about 900 $$^{circ}$$C in the IHX is provided to the hydrogen production facility such as IS system which produced the hydrogen by the thermo-chemical water-splitting iodine-sulfur process. The calculation to obtain a precise temperature distribution inside the IHX is required to the reliable design in the VHTR system with the design lifetime of 60 years. The 30 days operation in the HTTR with the reactor outlet coolant temperature of 850 $$^{circ}$$C has been performed and the temperature data for the IHX was obtained. The temperature calculation was performed to simulate the temperature distribution inside the IHX during the rated operation of the HTTR. The calculation result shows a good agreement with the experimental data and this calculation code was validated. It was confirmed that the IHX temperature calculation code was able to simulate precisely the temperature distribution inside the heat exchanger.

Journal Articles

Analysis on temperature distribution of reactor vessel cooling system during loss of coolant flow in HTTR

Takeda, Tetsuaki*; Ichimiya, Koichi*; Nishio, Hitoshi*; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi

Proceedings of 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7) (CD-ROM), 11 Pages, 2008/10

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 partial loss of coolant flow test was performed under the condition of the ATWS (Anticipated Transient without Scram). We are planning to perform the test of loss of coolant flow and stopping the vessel cooling system (VCS). The test of the loss of coolant flow as one of safety demonstration tests is carried out by tripping all gas circulators, and the stopping VCS test is performed continuously after the loss of coolant flow. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the VCS during the tests. It is necessary to consider the effect of thermal radiation from the RPV for evaluation of temperature of the VCS and concrete vessel.

Journal Articles

A Study of air ingress and its prevention in HTGR

Yan, X.; Takeda, Tetsuaki; Nishihara, Tetsuo; Ohashi, Kazutaka; Kunitomi, Kazuhiko; Tsuji, Nobumasa*

Nuclear Technology, 163(3), p.401 - 415, 2008/09

 Times Cited Count:10 Percentile:59.39(Nuclear Science & Technology)

A rupture of primary piping in HTGR represents a design basis event. In such a loss of coolant event a safety issue remains graphite oxidation damage to fuel and core should major air ingress take place through the breached primary boundary. The present study deals with the two most probable cases of air ingress. The first results from rupture of a standpipe. A design change proposed in the vessel top structure intends to rule out any probability of a standpipe rupture. The feasibility of the modified structure is evaluated. The second case results from rupture of a main coolant pipe. Experiment and analysis are performed to gain understanding of the multi-phased air ingress phenomena and accordingly a new mechanism of sustained counter-air diffusion is proposed that is fully passive and effective in preventing major air ingress in the event of main coolant pipe rupture. The results of the present study may lead to improved safety and economic design of the HTGR.

Journal Articles

The Preliminary analysis of the loss of primary coolant flow test in the HTTR

Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Goto, Minoru; Takeda, Tetsuaki*; Nakao, Yasuyuki*

Proceedings of 4th International Topical Meeting on High Temperature Reactor Technology (HTR 2008) (CD-ROM), 8 Pages, 2008/09

Loss of primary coolant flow test is under planning by using the HTTR to demonstrate the inherent safety features during the accident condition such as the depressurization accident which is selected as the severest accident in the HTGR. All the gas circulators are tripped in the test and the position of all control rods keeps its initial one. Because the core temperature increases just after the loss of coolant flow, the reactor power decreases according to coolant flow decrease due to negative reactivity feedback effect and the reactor becomes subcritical. The reactor performance after becoming subcritical during the loss of coolant flow is subject to a reactivity balance of core temperature and xenon concentration changes. The loss of primary coolant flow test in the HTTR simulates the depressurization accident and the data obtained from the test is useful for the validation and improvement of the calculation code applied to the safety analysis in the future HTGR.

Journal Articles

Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the high temperature engineering test reactor

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Journal of Power and Energy Systems (Internet), 2(2), p.790 - 803, 2008/00

The reactivity insertion 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, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. 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 effect of the model is formulated quantitatively with our proposed equation.

Journal Articles

Improvement of analysis technology for high temperature gas-cooled reactor by using data obtained in high temperature engineering test reactor

Nakagawa, Shigeaki; Tochio, Daisuke; Takamatsu, Kuniyoshi; Goto, Minoru; Takeda, Tetsuaki

Journal of Power and Energy Systems (Internet), 2(1), p.83 - 91, 2008/00

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 30 MW at a reactor outlet coolant temperature of about 950 $$^{circ}$$C 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.

Journal Articles

Application effect of region temperature coefficients and improvement of heat transfer analysis model in HTGR

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 6(3), p.262 - 275, 2007/09

The HTTR, which has thermal output of 30 MW, coolant inlet temperature of 395 $$^{circ}$$C and coolant outlet temperature of 850 $$^{circ}$$C/950 $$^{circ}$$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 HTGRs. The coolant flow reduction test demonstrates that rapid decrease of reactor power during tripping of the gas circulators is restrained by only the negative reactivity feedback effect without 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 a connection between some fuel channel models and whole core component model is adopted for calculating heat transfer in the core.

Journal Articles

Core dynamics analysis on reactivity insertion and loss of coolant flow tests for HTGRs

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 13 Pages, 2007/09

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). 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. This paper describes the validation results for the newly developed code using the experimental results of the safety demonstration tests. Especially, the reactivity was clarified using an original mathematical expression which shows the relationship among region temperature coefficient, region temperature rise and power distribution.

Journal Articles

Development of analytical methodology regarding reactor performance and safety characteristics of HTGR; Loss of coolant flow tests

Takamatsu, Kuniyoshi; Takeda, Tetsuaki; Nakagawa, Shigeaki; Goto, Minoru

Dai-12-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.213 - 214, 2007/06

Safety demonstration tests using the HTTR are in progress to verify the inherent safety features and to improve the safety technology and design methodology of high temperature gas-cooled reactors (HTGRs). The numerical analysis code was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel flow channels and twenty temperature coefficients in the core. This paper describes an analytical result of the loss of partial coolant flow test using the newly developed code. The analytical result of transient reactor power shows good agreement with the measured value during the test. Moreover, this paper refers to an analytical result of the loss of coolant flow test. The reactor power decreases to decay heat level due to the negative reactivity feedback effect of the core. Although the reactor power becomes critical again later, the peak power value is very small.

Journal Articles

Japanese technology on high-temperature gas-cooled reactor leading to world standard

Ogawa, Masuro; Hino, Ryutaro; Kunitomi, Kazuhiko; Onuki, Kaoru; Inagaki, Yoshiyuki; Takeda, Tetsuaki; Sawa, Kazuhiro

Genshiryoku eye, 53(4), p.26 - 33, 2007/04

no abstracts in English

Journal Articles

Applicability of heat transfer enhancement method using porous material to nuclear hydrogen production system

Takeda, Tetsuaki; Ichimiya, Koichi*

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 4 Pages, 2007/04

As for the development of the coupling technology between the HTGR and the hydrogen production system, JAEA have carried out the hydrogen production test with the steam reforming process by natural gas. In the HTGR hydrogen production system, disk type fins are attached on the outside surface of the catalyst tube and the tube is inserted into the guide tube to increase an amount of transferred heat in the present design of the steam reformer. However, we have to take the deterioration of the structure strength by attaching the fins and processing the tube surface into consideration with the design of the steam reformer. The objectives of this study are to develop a method for heat transfer enhancement using a porous material and to discuss the applicability of this method into the steam reformer of the nuclear hydrogen production system. An experiment has been performed using the simulated apparatus of the steam reformer to obtain the heat transfer and fluid flow characteristics.

Journal Articles

Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04

The numerical analysis code, ACCORD, has modified to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core structural materials for calculating the heat conduction between the fuel channels and the core in the case of the coolant flow reduction test. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the coolant flow reduction test by tripping one or two out of three gas circulators. Finally, the pre-analytical result of the coolant flow reduction test by tripping all gas circulators is also discussed. The reactor power decreases to decay heat level from 30 MW due to the negative reactivity feedback effect. Although the reactor power becomes critical again about five hours later, the peak power value is merely 2 MW.

Journal Articles

Improvement of analysis technologies for HTGR by using the HTTR data

Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Goto, Minoru; Takeda, Tetsuaki; Iyoku, Tatsuo

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04

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$$^{circ}$$C 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.

Journal Articles

Validation of neutronics calculation codes for VHTR nuclear design using HTTR experimental data

Goto, Minoru; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Takeda, Tetsuaki

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04

In order to validate applicability of neutronics calculation codes for a nuclear design of the VHTR, analysis of the core characteristics were performed for the HTTR. Additionally, effects of difference of nuclear data libraries on the core calculations for the HTTR were studied for JENDL (Japan), ENDF/B (U.S.A.) and JEFF (Europe). The calculation result of the HTTR excess reactivity at room temperature condition by the MVP was in good agreement with the experimental data within 0.4%$$Delta$$k/k and that by the SRAC, meanwhile, overestimated the experimental data about 1.5%$$Delta$$k/k. In consequence of the comparison between the HTTR core calculation results performed by the MVP and the experimental data, JENDL-3.3, ENDF/B-VI.8 and JEFF-3.1 yielded the excess reactivity agreement with the experiments within 0.4%$$Delta$$k/k, 0.7%$$Delta$$k/k and 0.7%$$Delta$$k/k, respectively.

Journal Articles

Research and development on system integration technology for connection of hydrogen production system to an HTGR

Inagaki, Yoshiyuki; Ohashi, Hirofumi; Inaba, Yoshitomo; Sato, Hiroyuki; Nishihara, Tetsuo; Takeda, Tetsuaki; Hayashi, Koji; Ogawa, Masuro

Nuclear Technology, 157(2), p.111 - 119, 2007/02

 Times Cited Count:10 Percentile:57.83(Nuclear Science & Technology)

Japan Atomic Energy Agency (JAEA) has been promoting research and development on the hydrogen production technology with a high temperature gas-cooled reactor (HTGR) with a view to contributing to the global warming issue and hydrogen energy society in the near future. The system integration technology for connection of the hydrogen production system to HTGR is one of the key technologies to put hydrogen production with nuclear energy to commercial use. Research and development on the system integration technology have been carried out about four items, that is, (a) control technology to keep reactor operation against thermal disturbance caused by the hydrogen production system, (b) estimation of tritium permeation from reactor to hydrogen, (c) countermeasure against explosion and (d) development of high temperature valve to isolate reactor and hydrogen production systems in accidents. This report describes research activities on the system integration technology at JAEA.

Journal Articles

Numerical analysis on air ingress behavior in GTHTR300-cogeneration system

Takeda, Tetsuaki; Yan, X.; Kunitomi, Kazuhiko

Journal of Power and Energy Systems (Internet), 1(1), p.24 - 35, 2007/00

The objective of this study is to clarify safety characteristics of a High Temperature Gas-Cooled Reactor for the pipe rupture accident. Japan Atomic Energy Agency (JAEA) has been developing the analytical code for the safety characteristics of the HTGR and carrying out design study of the gas turbine high temperature reactor of 300MWe nominal-capacity for hydrogen production, the GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). A numerical analysis of heat and mass transfer fluid flow with multi-component gas mixture has been performed to obtain the variation of the density of the gas mixture, and the onset time of natural circulation of air. From the results obtained in this analysis, it was found that the duration time of the air ingress by molecular diffusion would increase due to the existence of the recuperator in the GTHTR300C system.

Journal Articles

Thermal performance of a circular tube filled with a high porous material

Ichimiya, Koichi*; Takeda, Tetsuaki; Uemura, Takuya*; Norikuni, Tetsuya*

Nihon Kikai Gakkai Rombunshu, B, 72(723), p.2747 - 2752, 2006/11

no abstracts in English

129 (Records 1-20 displayed on this page)