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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

Temperature measurement of control rod using melt wire in High-Temperature Engineering Test Reactor (HTTR)

Hamamoto, Shimpei; Tochio, Daisuke; Ishii, Toshiaki; Sawahata, Hiroaki

Nippon Genshiryoku Gakkai Wabun Rombunshi, 16(4), p.169 - 172, 2017/12

A melt wire was installed at the tip of the control rod in order to measure the temperature of High Temperature engineering Test Reactor (HTTR). After experience with reactor scrum from the state of reactor power 100%, the melt wire was taken out from the control rod and appearance has been observed visually. It was confirmed that the melt wires with a melting point of 505 $$^{circ}$$C or less were melted, and the melt wires with a melting point of 651 $$^{circ}$$C or more were not melted. Therefore, it was found that the highest arrival temperature of tip of the control rods where the melt wires are installed reaches within the range of 505 to 651 $$^{circ}$$C. And it was found that the control rod temperature at the time of reactor scram does not exceed the using temperature criteria (900 $$^{circ}$$C) of Alloy 800H of the control rod sleeve.

JAEA Reports

Development of temperature measurement technology for control rod using melt wire in High Temperature engineering Test Reactor (HTTR)

Hamamoto, Shimpei; Sawahata, Hiroaki; Suzuki, Hisashi; Ishii, Toshiaki; Yanagida, Yoshinori

JAEA-Technology 2017-012, 20 Pages, 2017/06

JAEA-Technology-2017-012.pdf:7.9MB

A melt wire was installed at the tip of the control rod in order to measure the temperature of High Temperature engineering Test Reactor (HTTR). After experience with reactor scram from the state of reactor power 100%, the melt wire was taken out from the control rod and appearance has been observed visually. In this study, an exclusive device for taking out the melt wire was prepared. The take-out device functions as expected, and the melt wire was safely and reliably taken out using a remote manipulator. And because the visual observation of the melt wire was clearly carried out, we were successful in developing the control rod temperature measurement technology. It was confirmed that the melt wires with a melting point of 505$$^{circ}$$C or less were melted, and the melt wires with a melting point of 651$$^{circ}$$C or more were not melted. Therefore, it was found that the highest arrival temperature of tip of the control rods where the melt wires are installed reaches within the range of 505 to 651$$^{circ}$$C. And it was found that the control rod temperature at the time of reactor scram does not exceed the using temperature criteria (900$$^{circ}$$C) of Alloy 800H of the control rod sleeve.

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.

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

 Times Cited Count:0 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.

JAEA Reports

Shielding calculation by PHITS code during replacement works of startup neutron sources for HTTR operation

Shinohara, Masanori; Ishitsuka, Etsuo; Shimazaki, Yosuke; Sawahata, Hiroaki

JAEA-Technology 2016-033, 65 Pages, 2017/01

JAEA-Technology-2016-033.pdf:11.14MB

To reduce the neutron exposure dose for workers during the replacement works of the startup neutron sources of the High Temperature Engineering Test Reactor, calculations of the exposure dose in case of temporary neutron shielding at the bottom of fuels handling machine were carried out by the PHITS code. As a result, it is clear that the dose equivalent rate due to neutron radiation can be reduced to about an order of magnitude by setting a temporary neutron shielding at the bottom of shielding cask for the fuel handling machine. In the actual replacement works, by setting temporary neutron shielding, it was achieved that the cumulative equivalent dose of the workers was reduced to 0.3 man mSv which is less than half of cumulative equivalent dose for the previous replacement works; 0.7 man mSv.

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

 Times Cited Count:0 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

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.

Journal Articles

Establishment of integrity evaluation method for Reserved Shutdown System of High-Temperature engineering Test Reactor (HTTR)

Hamamoto, Shimpei; Kawamoto, Taiki; Kondo, Makoto; Sawahata, Hiroaki; Takada, Shoji; Shinozaki, Masayuki

Nippon Genshiryoku Gakkai Wabun Rombunshi, 15(2), p.66 - 69, 2016/06

High Temperature engineering Test Reactor (HTTR) has the reactivity control system which is accompanied with the Reserved Shutdown System (RSS). The RSS consists of B$$_{4}$$C/C pellets, guide tube, electric plug, motor which contains brake and reducer, and so on. In accidents when the control rods cannot be inserted, an electric plug is pulled out by motor and the B$$_{4}$$C/C pellets fall into the core by gravity. It was revealed that the motor in the RSS drive mechanism did not work as the result of pre-start up checks as described followings: (1) The oil which was separated from a grease of motor reducer flowed down from gap of oil seal, (2) the separated oil penetrated into the brake, (3) the penetrated oil was mixed with abrasion powder released from brake disk, finally, (4) the adhesive mixture blocked the rotation of the motor. A new evaluation method was proposed to detect a sign of the motor sticking. Through the overhaul inspection of all RSS drive mechanisms of HTTR, it was revealed that the proposed method was effective to evaluate the integrity of the RSS drive mechanism.

Journal Articles

Improvement of exchanging method of neutron startup source of high temperature engineering test reactor

Sawahata, Hiroaki; Shimazaki, Yosuke; Ishitsuka, Etsuo; Yamazaki, Kazunori; Yanagida, Yoshinori; Fujiwara, Yusuke; Takada, Shoji; Shinozaki, Masayuki; Hamamoto, Shimpei; Tochio, Daisuke

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 8 Pages, 2016/06

In the HTTR, $$^{252}$$Cf is loaded in the reactor core as a neutron startup source and changed at frequency. In this exchange work, there were two technical issues; slightly higher radiation exposure of workers by neutron leakage and reliability of neutron source transportation container in handling. To reduce the radiation dose by neutron leakage, detail numerical evaluations using PHITS code were carried out, the effective shielding method for fuel handling machine was proposed. Easily removable polyethylene blocks and particles were used as the neutron shielding, and installed in the cooling paths of the fuel handling machine. As a result, the collective effective dose by neutron was reduced from about 700 man-microSv to about 300 man-microSv. As to the neutron source transportation container, the handling performance was improved and the handling work was safety accomplished by downsizing.

Journal Articles

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

Shimazaki, Yosuke; Ono, Masato; Tochio, Daisuke; Takada, Shoji; Sawahata, Hiroaki; Kawamoto, Taiki; Hamamoto, Shimpei; Shinohara, Masanori

Proceedings of International Topical Meeting on Research Reactor Fuel Management and Meeting of the International Group on Reactor Research (RRFM/IGORR 2016) (Internet), p.1034 - 1042, 2016/03

In High Temperature Engineering Test Reactor (HTTR), three neutron holders containing $$^{252}$$Cf with 3.7 GBq for each are loaded in the graphite blocks and inserted into the reactor core as a neutron startup source which is changed at the interval of approximately ten years. These neutron holders containing the neutron sources are transported from the dealer's hot cell to HTTR using the transportation container. The holders loading to the graphite block are carried out in the fuel handling machine maintenance pit of HTTR. There were two technical issues for the safety handling work of the neutron holder. The one is the radiation exposure caused by significant movement of the container due to an earthquake, because the conventional transportation container was so large ($$phi$$1240 mm, h1855 mm) that it can not be fixed on the top floor of maintenance pit by bolts. The other is the falling of the neutron holder caused by the difficult remote handling work, because the neutron holder capsule was also so long ($$phi$$155 mm, h1285 mm) that it can not be pulled into the adequate working space in the maintenance pit. Therefore, a new and low cost transportation container, which can solve the issues, was developed. To avoid the neutron and $$gamma$$ ray exposure, smaller transportation container ($$phi$$820mm, h1150 mm) which can be fixed on the top floor of maintenance pit by bolts was developed. In addition, to avoid the falling of the neutron holder, smaller neutron holder capsule ($$phi$$75 mm, h135 mm) with simple handling mechanism which can be treated easily by manipulator was also developed. As the result of development, the neutron holder handling work was safely accomplished. Moreover, a cost reduction for manufacturing was also achieved by simplifying the mechanism of neutron holder capsule and downsizing.

Journal Articles

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

Honda, Yuki; Fujimoto, Nozomu; Sawahata, Hiroaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

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

Improvement of cell model for control rod in reflector region of high temperature test engineering reactor

Honda, Yuki; Fujimoto, Nozomu; Sawahata, Hiroaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

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

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

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

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.

Journal Articles

Development of the maintenance technologies for the future high-temperature gas cooled reactor (HTGR) using operating experiences acquired in high-temperature engineering test reactor (HTTR)

Shimazaki, Yosuke; Homma, Fumitaka; Sawahata, Hiroaki; Furusawa, Takayuki; Kondo, Masaaki

Journal of Nuclear Science and Technology, 51(11-12), p.1413 - 1426, 2014/11

 Times Cited Count:1 Percentile:87.72(Nuclear Science & Technology)

Journal Articles

Development of operation and maintenance technology for HTGRs by using HTTR (High Temperature engineering Test Reactor)

Shimizu, Atsushi; Kawamoto, Taiki; Tochio, Daisuke; Saito, Kenji; Sawahata, Hiroaki; Homma, Fumitaka; Furusawa, Takayuki; Saikusa, Akio; Takada, Shoji; Shinozaki, Masayuki

Nuclear Engineering and Design, 271, p.499 - 504, 2014/05

 Times Cited Count:4 Percentile:58.26(Nuclear Science & Technology)

The long term high temperature operation using HTTR was carried out to establish the technical basis of HTGR in the high temperature test operation mode during 50-day since January till March, 2010. It is necessary to demonstrate the stability of plant during long-term operation in order to attain the stable supply of the high temperature heat to the planned heat utilization system of HTTR. Test data obtained in the operation were evaluated for the technical issues which were extracted before the operation. As the results, Stability and reliability of the components and facility was demonstrated by evaluating the heat transfer performance of high temperature components, the performance of pressure control to compensate helium gas leak, the reliability of the dynamic components such as helium gas circulators, the performance of heat-up protection of radiation shielding. Through the operation, the technical basis for the operation and maintenance technology of HTGRs was established.

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