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Journal Articles

Establishment of numerical model to investigate heat transfer and flow characteristics by using scale model of vessel cooling system for HTTR

Takada, Shoji; Narayana, I. W.*; Nakatsuru, Yukihiro*; Terada, Atsuhiko; Murakami, Kenta*; Sawa, kazuhiko*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 13 Pages, 2019/05

In the loss of core cooling test using HTTR, a technical issue is to improve prediction accuracy of temperature distribution of components in vessel cooling system (VCS). An establishment of reasonable 2D model was started by using numerical code FLUENT, which was validated using the test data by 1/6 scale model of VCS for HTTR. The pressure vessel (PV) temperature was set around 200$$^{circ}$$C attributed to relatively high ratio of natural convection heat transfer around 20% in total heat removal, which is useful for code to experiment benchmark to improve prediction accuracy. It is necessary to confirm heat transfer flow characteristics around the top of PV which is heated up by natural convection flow which was considered to be affected by separation, re-adhesion and transition flow. The k-$$omega$$-SST model was selected for turbulent calculation attributed to predict the effects mentioned above adequately. The numerical results using the k-$$omega$$-SST model reproduced the temperature distribution of PV especially the top region which is considered to be affected by separation, re-adhesion and transition flow in contract to that using k-$$varepsilon$$ model which does not account the effects.

Journal Articles

Feasibility study of new applications at the high-temperature gas-cooled reactor

Ho, H. Q.; Honda, Yuki*; Hamamoto, Shimpei; Ishii, Toshiaki; Takada, Shoji; Fujimoto, Nozomu*; Ishitsuka, Etsuo

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

Journal Articles

Numerical evaluation on fluctuation absorption characteristics based on nuclear heat supply fluctuation test using HTTR

Takada, Shoji; Honda, Uki*; Inaba, Yoshitomo; Sekita, Kenji; Nemoto, Takahiro; Tochio, Daisuke; Ishii, Toshiaki; Sato, Hiroyuki; Nakagawa, Shigeaki; Sawa, Kazuhiro*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

Nuclear heat utilization systems connected to HTGRs will be designed on the basis of non-nuclear grade standards for easy entry of chemical plant companies, requiring reactor operations to continue even if abnormal events occur in the systems. The inventory control is considered as one of candidate methods to control reactor power for load following operation for siting close to demand area, in which the primary gas pressure is varied while keeping the reactor inlet and outlet coolant temperatures constant. Numerical investigation was carried out based on the results of nuclear heat supply fluctuation tests using HTTR by non-nuclear heating operation to focus on the temperature transient of the reactor core bottom structure by imposing stepwise fluctuation on the reactor inlet temperature under different primary gas pressures below 120C. As a result, it was emerged that the fluctuation absorption characteristics are not deteriorated by lowering pressure. It was also emerged that the reactor outlet temperature did not reach the scram level by increasing the reactor inlet temperature 10 C stepwise at 80% of the rated power as same with the full power case.

Journal Articles

Comprehensive seismic evaluation of HTTR against the 2011 off the Pacific coast of Tohoku Earthquake

Ono, Masato; Iigaki, Kazuhiko; Sawahata, Hiroaki; Shimazaki, Yosuke; Shimizu, Atsushi; Inoi, Hiroyuki; Kondo, Toshinari; Kojima, Keidai; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 4(2), p.020906_1 - 020906_8, 2018/04

On March 11th, 2011, the 2011 off the Pacific coast of Tohoku Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the High Temperature Engineering Test Reactor (HTTR) had been stopped under the periodic inspection and maintenance of equipment and instruments. A comprehensive integrity evaluation was carried out for the HTTR facility because the maximum seismic acceleration observed at the HTTR exceeded the maximum value of design basis earthquake. The concept of comprehensive integrity evaluation is divided into two parts. One is the "visual inspection of equipment and instruments". The other is the "seismic response analysis" for the building structure, equipment and instruments using the observed earthquake. All equipment and instruments related to operation were inspected in the basic inspection. The integrity of the facilities was confirmed by comparing the inspection results or the numerical results with their evaluation criteria. As the results of inspection of equipment and instruments associated with the seismic response analysis, it was judged that there was no problem for operation of the reactor, because there was no damage and performance deterioration. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013 and 2015. Additionally, the integrity of control rod guide blocks was also confirmed visually when three control rod guide blocks and six replaceable reflector blocks were taken out from reactor core in order to change neutron startup sources in 2015.

Journal Articles

Investigation of uncertainty caused by random arrangement of coated fuel particles in HTTR criticality calculations

Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji

Annals of Nuclear Energy, 112, p.42 - 47, 2018/02

 Times Cited Count:1 Percentile:100(Nuclear Science & Technology)

Journal Articles

Loss of core cooling test with one cooling line inactive in Vessel Cooling System of High-Temperature Engineering Test Reactor

Fujiwara, Yusuke; Nemoto, Takahiro; Tochio, Daisuke; Shinohara, Masanori; Ono, Masato; Takada, Shoji

Journal of Nuclear Engineering and Radiation Science, 3(4), p.041013_1 - 041013_8, 2017/10

In HTTR, the test was carried out at the reactor thermal power of 9 MW under the condition that one cooling line of VCS was stopped to simulate the partial loss of cooling function from the surface of RPV in addition to the loss of forced cooling flow in the core simulation. The test results showed that temperature change of the core internal structures and the biological shielding concrete was slow during the test. Temperature of RPV decreased several degrees during the test. The temperature decrease of biological shielding made of concrete was within 1$$^{circ}$$C. The numerical result simulating the detail configuration of the cooling tubes of VCS showed that the temperature rise of cooling tubes of VCS was about 15$$^{circ}$$C, which is sufficiently small, which did not significantly affect the temperature of biological shielding concrete. As the results, it was confirmed that the cooling ability of VCS can be kept in case that one cooling line of VCS is lost.

Journal Articles

Numerical investigation of the random arrangement effect of coated fuel particles on the criticality of HTTR fuel compact using MCNP6

Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji

Annals of Nuclear Energy, 103, p.114 - 121, 2017/05

 Times Cited Count:3 Percentile:42.02(Nuclear Science & Technology)

Journal Articles

Benchmark study on realized random packing model for coated fuel particles of HTTR using MCNP6

Ho, H. Q.; Morita, Keisuke*; Honda, Yuki; Fujimoto, Nozomu*; Takada, Shoji

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04

Journal Articles

Measurement of temperature response of intermediate heat exchanger in heat application system abnormal simulating test using HTTR

Ono, Masato; Fujiwara, Yusuke; Honda, Yuki; Sato, Hiroyuki; Shimazaki, Yosuke; Tochio, Daisuke; Homma, Fumitaka; Sawahata, Hiroaki; Iigaki, Kazuhiko; Takada, Shoji

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 5 Pages, 2017/04

Japan Atomic Energy Agency (JAEA) has carried out research and developments towards nuclear heat utilization of High Temperature Gas-cooled Reactor (HTGR) using High Temperature Engineering Test Reactor (HTTR). The nuclear heat utilization systems connected to HTGR will be designed on the basis of non-nuclear-grade standards in terms of easier entry for the chemical plant companies and the construction economics of the systems. Therefore, it is necessary that the reactor operations continue even if abnormal events occur in the systems. Heat application system abnormal simulating test with HTTR was carried out in non-nuclear heating operation to focus on the thermal effect in order to obtain data of the transient temperature behavior of the metallic components in the Intermediate Heat Exchanger (IHX). The IHX is the key components to connect the HTTR with the heat application system. In the test, the coolant helium gas temperature was heated up to 120$$^{circ}$$C by the compression heat of the gas circulators in the HTTR under the ideal condition to focus on the heat transfer. The tests were conducted by decreasing the helium gas temperature stepwise by increasing the mass flow rate to the air cooler. The temperature responses of the IHX were investigated. For the components such as the heat transfer tubes and heat transfer enhancement plates of IHX, the temperature response was slower in the lower position in comparison with the higher position. The reason is considered that thermal load fluctuation is imposed in the secondary helium gas which flows from the top to the bottom in the heat transfer tubes of the IHX. The test data are useful to verify the numerical model of the safety evaluation code.

Journal Articles

Study of the reduction method of the helium gas leakage from bolted gasket flanged connection for HTGRs

Hamamoto, Shimpei; Takada, Shoji

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 4 Pages, 2017/04

JAEA Reports

Neutronic characteristic of HTTR fuel compact with various packing models of coated fuel particle

Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji

JAEA-Technology 2016-040, 16 Pages, 2017/03

JAEA-Technology-2016-040.pdf:2.89MB

To study the packing effects of the truncated coated fuel particle on the criticality for the High Temperature engineering Test Reactor (HTTR), four alternative models including the truncated uniform model, the non-truncated uniform model, the truncated random model and the non-truncated random model for the arrangement of CFP in fuel compact were used, and the neutronic and criticality calculation were performed by using Monte Carlo MCNP6 code with ENDF/B-VII.1 cross section library. The results showed that the infinite multiplication factors (k$$_{rm inf}$$) in the truncated models were lower than those of the non-truncated models regardless of the uniform or random arrangement, and the four factors in the four-factor-formula showed that the difference of k$$_{rm inf}$$ was mainly attributed to the resonance escape probability. The difference in resonance escape probability is caused by the increase of capture reactions in the resonance region as the influence of spatial-self-shielding-effect. It is because the equivalent kernel diameter of the CFP for the truncated model is smaller than that of the non-truncated model.

JAEA Reports

Development of transportation container for neutron startup source of High Temperature Engineering Test Reactor (HTTR)

Shimazaki, Yosuke; Sawahata, Hiroaki; Yanagida, Yoshinori; Shinohara, Masanori; Kawamoto, Taiki; Takada, Shoji

JAEA-Technology 2016-038, 36 Pages, 2017/02

JAEA-Technology-2016-038.pdf:8.75MB

The High Temperature Engineering Test Reactor (HTTR) has three neutron startup sources (NSs) in the reactor core, each of which consists of $$^{252}$$Cf with 3.7GBq The NSs are exchanged at the interval of approximately 7 years. The NS holders including NSs are transported from the dealer's hot cell to the reactor facility of HTTR using a transportation container. The loading work of NS holders to the Control Rod guide blocks is subsequently carried out in the fuel handling machine maintenance pit of HTTR. Following technical issues were extracted from the experiences in the past two exchange works of NSs to develop a safety handling procedure; (1) The reduction and prevention of radiation exposure of workers. (2) The exclusion of falling of NS holder. Then, a new transportation container special to the NSs of HTTR was developed to solve the technical issues while keeping the cost as low as that for overhaul of conventional container and satisfying the regulation of A type transportation package.

Journal Articles

Improvement of neutron startup source handling work by developing new transportation container for High-Temperature engineering Test Reactor (HTTR)

Shimazaki, Yosuke; Sawahata, Hiroaki; Shinohara, Masanori; Yanagida, Yoshinori; Kawamoto, Taiki; Takada, Shoji

Journal of Nuclear Science and Technology, 54(2), p.260 - 266, 2017/02

 Percentile:100(Nuclear Science & Technology)

The High-Temperature engineering Test Reactor (HTTR) has three neutron startup sources (NSs) in the reactor core, each of which consists of $$^{252}$$Cf with 3.7 GBq and is contained in a small capsule, installed in NS holder and subsequently in a control guide block (CR block). The NSs are exchanged at the interval of approximately 7 years. The NS holders are transported from the dealer's hot cell to the reactor facility of HTTR using a transportation container. The loading work of NS holders to the CR blocks is subsequently carried out in the fuel handling machine maintenance pit of HTTR. Technical issues, which are the reduction and prevention of radiation exposure of workers and the exclusion of falling of NS holder, were extracted from the experiences in past two exchange works of NSs to develop a safety handling procedure. Then, a new transportation container special to the NSs of HTTR was developed to solve the technical issues while keeping the cost as low as that for overhaul of conventional container. As the results, the NS handling work using the new transportation container was safely accomplished by developing the new transportation container which can reduce the risks of radiation exposure dose of workers and exclude the falling of NS holder.

Journal Articles

Burn-up dependency of control rod position at zero-power criticality in the high-temperature engineering test reactor

Honda, Yuki; Fujimoto, Nozomu*; Sawahata, Hiroaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 3(1), p.011013_1 - 011013_4, 2017/01

The operating data of the HTTR with burn-up is very important for developments of HTGRs. There is a temperature distribution in a core in full power operation. The temperature distribution in a core makes it difficult to validate of calculation code. On the other hands, the data of the control rod position at criticality at zero power have been measured at the beginning of each operation cycle. The temperature distribution in a core at zero power is uniform. Therefore, the data at zero power are suitable for validation of calculation code. In this study, the calculated control rod positions at zero power criticality with burn-up are compared with the experimental data with correlation of core temperature. The calculated results of criticality control rod position at zero power show good agreement to the experimental data. It means that calculated result shows appropriate decrease in uranium and accumulation in plutonium decrease in burnable absorber with burn-up.

Journal Articles

Study on sensitivity of control rod cell model in reflector region of high-temperature engineering test reactor

Honda, Yuki; Fujimoto, Nozomu*; Sawahata, Hiroaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 3(1), p.011005_1 - 011005_6, 2017/01

In the HTTR, a two-step control rods insertion method for reactor scram is adopted. In the method, control rods at reflector region are inserted at the scram signal is initiated. The core should keep its subcriticality by reflector region control rods. Therefore, precise evaluation of control rods reactivity worth for reflector region is necessary. However, all cross section of control rods has been prepared for control rod in fuel region because the reactivity value of control rods in the fuel region is larger than that of control rods in the reflector region. This paper proposed the revised method of preparing the control rod cross section for first step control rod in reflector region.

Journal Articles

Investigation of absorption characteristics for thermal-load fluctuation using HTTR

Tochio, Daisuke; Honda, Yuki; Sato, Hiroyuki; Sekita, Kenji; Homma, Fumitaka; Sawahata, Hiroaki; Takada, Shoji; Nakagawa, Shigeaki

Journal of Nuclear Science and Technology, 54(1), p.13 - 21, 2017/01

 Percentile:100(Nuclear Science & Technology)

GTHTR300C is designed and developed in JAEA. 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. Then, it is necessary to demonstrate that the thermal-load fluctuation should be absorbed by the reactor system so as to continue the stable and safety operation could be continued. 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 IHX. As the result it was revealed that the reactor has the larger absorption capacity of thermal-load fluctuation than expected one, and 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 there result that the reactor and the IHX has effective absorption capacity of the thermal-load fluctuation generated in the heat utilization system. Moreover it was confirmed that the safety estimation code based on RELAP5/MOD3 can represents the thermal-load fluctuation absorption behavior conservatively.

Journal Articles

Investigation of countermeasure against local temperature rise in vessel cooling system in loss of core cooling test without nuclear heating

Ono, Masato; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.044502_1 - 044502_4, 2016/10

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to verify safety evaluation codes to investigate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from gas circulators with stopping water flow in the VCS, the local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1$$^{circ}$$C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

Journal Articles

Nuclear heat supply fluctuation tests by non-nuclear heating with HTTR

Inaba, Yoshitomo; Sekita, Kenji; Nemoto, Takahiro; Honda, Yuki; Tochio, Daisuke; Sato, Hiroyuki; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.041001_1 - 041001_7, 2016/10

The nuclear heat utilization systems connected to High Temperature Gas-cooled Reactors (HTGRs) will be designed on the basis of non-nuclear grade standards in terms of the easier entry of chemical plant companies and the construction economics of the systems. Therefore, it is necessary that the reactor operations can be continued even if abnormal events occur in the systems. The Japan Atomic Energy Agency has developed a calculation code to evaluate the absorption of thermal load fluctuations by the reactors when the reactor operations are continued after such events, and has improved the code based on the High Temperature engineering Test Reactor (HTTR) operating data. However, there were insufficient data on the transient temperature behavior of the metallic core side components and the graphite core support structures corresponding to the fluctuation of the reactor inlet coolant temperature for further improvement of the code. Thus, nuclear heat supply fluctuation tests with the HTTR were carried out in non-nuclear heating operation to focus on thermal effect. In the tests, the coolant helium gas temperature was heated up to 120$$^{circ}$$C by the compression heat of the gas circulators in the HTTR, and a sufficiently high fluctuation of 17$$^{circ}$$C by devising a new test procedure was imposed on the reactor inlet coolant under the ideal condition without the effect of the nuclear power. Then, the temperature responses of the metallic core side components and the graphite core support structures were investigated. The test results adequately showed as predicted that the temperature responses of the metallic components are faster than those of the graphite structures, and the mechanism of the thermal load fluctuation absorption by the metallic components was clarified.

Journal Articles

Development of the prediction technology of cable disconnection of in-core neutron detector for the future high-temperature gas-cooled reactors

Shimazaki, Yosuke; Sawahata, Hiroaki; Kawamoto, Taiki; Suzuki, Hisashi; Shinohara, Masanori; Honda, Yuki; Katsuyama, Kozo; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.041008_1 - 041008_5, 2016/10

Maintenance technologies for the reactor system have been developed by using the high-temperature engineering test reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for the future HTGRs by shifting from the time-based maintenance to condition-based maintenance. The technical issue of the maintenance of in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because that it is difficult to observe directly inside of the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside of the reactor by using the time domain reflectometry and so on. The disconnection position, which was specified by the electrical method, was identified by non-destructive and destructive inspection. The accumulated data is expected to be contributed for advanced maintenance of future HTGRs.

JAEA Reports

HTTR thermal load fluctuation test (non-nuclear heating test); Confirmation of HTGR system response against temperature transient

Honda, Yuki; Tochio, Daisuke; Nakagawa, Shigeaki; Sekita, Kenji; Homma, Fumitaka; Sawahata, Hiroaki; Sato, Hiroyuki; Sakaba, Nariaki; Takada, Shoji

JAEA-Technology 2016-016, 16 Pages, 2016/08

JAEA-Technology-2016-016.pdf:2.84MB

A system analysis code is validated with the thermal-load fluctuation absorption test with nun-nuclear heating by using the High Temperature Engineering test Reactor (HTTR) to clarify the High Temperature Gas-cooled Reactor (HTGR) system response against temperature transient. The thermal-load fluctuation absorption test consists on the thermal load fluctuation tests (non-nuclear heating) and heat application system abnormal simulating test (non-nuclear heating). The HTGR reactor response against temperature transient is clarified in the thermal load fluctuation test (non-nuclear heating). The Intermediate Heat Exchanger (IHX) reactor response against temperature transient is clarified in the heat application system abnormal simulating test (non-nuclear heating). With the two HTTR non-nuclear heating test, HTGR system response against temperature transient is obtained.

128 (Records 1-20 displayed on this page)