Hamamoto, Shimpei; Shimizu, Atsushi; Inoi, Hiroyuki; Tochio, Daisuke; Homma, Fumitaka; Sawahata, Hiroaki; Sekita, Kenji; Watanabe, Shuji; Furusawa, Takayuki; Iigaki, Kazuhiko; et al.
Nuclear Engineering and Design, 388, p.111642_1 - 111642_11, 2022/03
Following the Fukushima Daiichi Nuclear Power Plant accident in 2011, the Japan Atomic Energy Agency adapted High-Temperature engineering Test Reactor (HTTR) to meet the new regulatory requirements that began in December 2013. The safety and seismic classifications of the existing structures, systems, and components were discussed to reflect insights regarding High Temperature Gas-cooled Reactors (HTGRs) that were acquired through various HTTR safety tests. Structures, systems, and components that are subject to protection have been defined, and countermeasures to manage internal and external hazards that affect safety functions have been strengthened. Additionally, measures are in place to control accidents that may cause large amounts of radioactive material to be released, as a beyond design based accident. The Nuclear Regulatory Commission rigorously and appropriately reviewed this approach for compliance with the new regulatory requirements. After nine amendments, the application to modify the HTTR's installation license that was submitted in November 2014 was approved in June 2020. This response shows that facilities can reasonably be designed to meet the enhanced regulatory requirements, if they reflect the characteristics of HTGRs. We believe that we have established a reference for future development of HTGR.
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 950C, 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.
Inaba, Yoshitomo; Hamamoto, Shimpei; Furusawa, Takayuki; Saikusa, Akio; Sakaba, Nariaki
Journal of Nuclear Science and Technology, 51(11-12), p.1373 - 1386, 2014/11
A main objective to install filters upstream of primary gas circulators in the high temperature engineering test reactor (HTTR), besides having a primary helium purification system, is the reduction and removal of circulating dust in the primary circuit. A problem encountered with the filters during the initial operations of the HTTR was that the differential pressure across the filters had increased excessively over the duration of the operations so that the differential pressure would be expected to exceed the limit value regulated in the HTTR operation manual. It was speculated that either the carbon traced back chemical reactions, the debris from mechanical contacts or both of these sources might be captured by the filters. Then, the filters were replaced and inspected to identify the cause of the increase of the filter differential pressure. As a result, it was found that the increase is caused by clogging of the filters by the dust traced back to the physical contact of the piston rings of the gas circulators equipped in the primary helium purification system. Hence, prismatic block-type very high temperature reactors (VHTRs) do not continuously supply carbon dust from the cores in operation.
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
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.
Shimizu, Atsushi; Kawamoto, Taiki; Tochio, Daisuke; Saito, Kenji; Sawahata, Hiroaki; Homma, Fumitaka; Furusawa, Takayuki; Saikusa, Akio; Shinozaki, Masayuki
Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 8 Pages, 2012/10
To establish the technical basis of HTGR, the long term high temperature operation using HTTR was carried out during 50-day in 2010. It is necessary to demonstrate the stability of plant during long-term operation and the reliability of components and facilities special to HTGRs, in order to attain the stable supply of the high temperature heat to the planned hydrogen production 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 special to HTGRs was demonstrated by evaluating the heat transfer performance of high temperature components, the helium gas leak tightness, 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 were established.
Tochio, Daisuke; Hamamoto, Shimpei; Inoi, Hiroyuki; Shimazaki, Yosuke; Sekita, Kenji; Kondo, Masaaki; Saikusa, Akio; Kameyama, Yasuhiko; Saito, Kenji; Emori, Koichi; et al.
JAEA-Technology 2010-038, 57 Pages, 2010/12
In HTTR, in-service operation is conducted through the rise-to power operation with rated operation or high-temperature test operation from achievement of first criticality at 1998. To make practical use HTGR system, it must be demonstrated to supply stable heat to heat utilization system for long-term. In HTTR, high-temperature/parallel-loaded long-term operation had been performed from January 2010. As the result, it was demonstrated to supply stable heat to heat utilization system for 50 days with HTTR, moreover, various long-term operation data were gained. This paper reports the characteristics of the high-temperature long-term operation for HTTR obtained from the operation.
Tochio, Daisuke; Nojiri, Naoki; Hamamoto, Shimpei; Inoi, Hiroyuki; Sekita, Kenji; Kondo, Masaaki; Saikusa, Akio; Kameyama, Yasuhiko; Saito, Kenji; Fujimoto, Nozomu
JAEA-Technology 2009-005, 47 Pages, 2009/05
HTTR is now conducted in-service operation through the rise-to power operation with rated operation or high-temperature test operation from achievement of first criticality at 1998. In order to demonstrate to supply stable heat to heat utilization system for long-term, HTTR was conducted rated/parallel-loaded 30-days operation. This paper reports the characteristics of long-term operation for HTTR.
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 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 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.
Furusawa, Takayuki; Saikusa, Akio; Hamamoto, Shimpei; Nemoto, Takahiro; Shinohara, Masanori; Isozaki, Minoru
JAEA-Technology 2007-066, 38 Pages, 2008/01
In the HTTR rise-to-power test which was performed from April in 2000 as phase 1 up to 10MW, nitrogen gas remained in the air cooler which release the heat to atmosphere. This residence nitrogen gas causes the reduction of the thermal performance of the air cooler. So, it was impossible that heat generated reactor core could not remove when reactor operated full power operation. A mockup test was carried out to investigate the occurrence mechanism of the residence nitrogen gas. From a result of the mockup test, we clarified that the marked wave rise in the water pressurizer and the melting velocity of the nitrogen gas into the pressurized water is thought to be higher than expected. Therefore, we installed a hollow type plate, multi-hole type plate and so on in the water pressurizer. As a result, it was confirmed that no residence nitrogen gas in the air cooler during rise-to-power test and normal operation. Consequently, the hollow type plate and multi-hole type plate were effective for prevention of the residence nitrogen gas in the air cooler. This paper describes the results of the mockup test and the improvement of the water pressurizer.
Iigaki, Kazuhiko; Saikusa, Akio; Sawahata, Hiroaki; Shinozaki, Masayuki; Tochio, Daisuke; Homma, Fumitaka; Tachibana, Yukio; Iyoku, Tatsuo; Kawasaki, Kozo; Baba, Osamu*
JAEA-Review 2006-010, 90 Pages, 2006/07
Gas Cooled Reactor has long history of nuclear development, and High Temperature Gas Cooled Reactor (HTGR) has been expected that it can be supply high temperature energy to chemical industry and to power generation from the points of view of the safety, the efficiency, the environment and the economy. The HTGR design is tried to installed passive safety equipment. The current licensing review guideline was made for a Low Water Reactor (LWR) on safety evaluation therefore if it would be directly utilized in the HTGR it needs the special consideration for the HTGR. This paper describes that investigation result of the safety design and the safety evaluation traditions for the HTGR, comparison the safety design and safety evaluation feature for the HTGT with it's the LWR, and reflection for next HTGR based on HTTR operational experiment.
Tochio, Daisuke; Watanabe, Shuji; Saikusa, Akio; Oyama, Sunao; Nemoto, Takahiro; Hamamoto, Shimpei; Shinohara, Masanori; Isozaki, Minoru; Nakagawa, Shigeaki
JAEA-Technology 2006-005, 83 Pages, 2006/02
In High Temperature Engineering Test Reactor (HTTR), the rated thermal power of 30MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary cooling system, such as the intermediate heat exchanger (IHX) and the secondary pressurized water cooler (SPWC). The heat exchangers in the main cooling system are required the heat exchange performance to remove the reactor-generated-heat of 30MW under the condition of reactor coolant outlet temperature of 850 C/ 950 C. Therefore, the heat exchanges are required to satisfy the design criteria of heat exchange performance. In this report, heat exchange performance of the SPWC in the main cooling system was evaluated with the rise-to-power-up test and the in-service operation data. Moreover, evaluated value is compared with designed one, it is confirmed that the SPWC has required heat exchange performance.
Hamamoto, Shimpei; Saikusa, Akio; Shinohara, Masanori; Tachibana, Yukio
Koon Gakkai-Shi, 32(1), p.43 - 49, 2006/01
The intermediate heat exchanger (IHX) of the High Temperature Engineering Test Reactor (HTTR) is one of the high temperature components of the HTTR and a helium-helium type heat exchanger with the heat capacity of 10 MW. The internal structures such as heat transfer tubes made of Hastelloy XR are used normally at the temperature higher than 900C. They compose the pressure boundary between the primary and secondary helium. Their creep strain and creep damage are evaluated based on the high temperature structural design guideline. The IHX is the first high temperature heat exchanger applied to the reactor. Therefore, Research and development works on the IHX were carried out to confirm the structural integrity of the IHX elements as follows. Experimental and analytical studies were carried out to confirm the structural integrity of the IHX as follows: (1) creep collapse of the tube against external pressure, (2) creep fatigue of the tube against thermal stress, (3) seismic behavior of the tube bundles, (4) thermal hydraulic behavior of the tube bundles and (5) in-service inspection technology of the tube. This report describes the objective and component tests procedure on the IHX and its results.
Fujimoto, Nozomu; Tachibana, Yukio; Saikusa, Akio*; Shinozaki, Masayuki; Isozaki, Minoru; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.273 - 281, 2004/10
From a viewpoint of heat leakage, there were two incidents during HTTR power-rise-tests. One was a temperature rise of the primary upper shielding, and the other was a temperature rise of the core support plate. Causes of the both incidents were small amount of helium flow in structures. For the temperature rise of the primary upper shielding, countermeasures to reduce the small amount of helium flow, enhancement of heat release and installation of thermal insulator were taken. For the temperature rise of the core support plate, temperature evaluations were carried out again considering the small amount of helium flow and design temperature of the core support plate was revised. By these countermeasures, the both temperatures were kept below their limits.
Tachibana, Yukio; Nakagawa, Shigeaki; Takeda, Takeshi; Saikusa, Akio; Furusawa, Takayuki; Takamatsu, Kuniyoshi; Sawa, Kazuhiro; Iyoku, Tatsuo
Nuclear Engineering and Design, 224(2), p.179 - 197, 2003/09
no abstracts in English
Saikusa, Akio*; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo
JAERI-Data/Code 2002-027, 34 Pages, 2003/02
High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28,1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is a first Reactor Cavity Cooling System applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it is confirmed that the VCS heat removal at 30 MW power operation is higher than 0.3MW. This paper shows outline of the VCS and test results on the VCS performance.
Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi; Nojiri, Naoki; Takeda, Takeshi; Saikusa, Akio; Ueta, Shohei; Kojima, Takao; Takada, Eiji*; Saito, Kenji; et al.
JAERI-Tech 2002-069, 87 Pages, 2002/08
Rise-to-power test in the HTTR has been performed from April 23rd to June 6th in 2000 as phase 1 test up to 10MW, from January 29th to March 1st in 2001 as phase 2 test up to 20MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20MW in the high temperature test operation mode. Phase 4 test to achieve the thermal reactor power of 30MW started from October 23rd in 2001. On December 7th it was confirmed that the thermal reactor power reached to 30MW and the reactor outlet coolant temperature reached to 850C. JAERI obtained the certificate of pre-operation test from MEXT because all the pre-operation tests by MEXT were passed successfully. From the test results of rise-up-power test up to 30MW, the performance of reactor and cooling system were confirmed, and it was confirmed that an operation of reactor facility could be performed safely. Some problems to be solved were found through tests. By means of solving them, the reactor operation with the reactor outlet coolant temperature of 950C will be achievable.
Tachibana, Yukio; Nakagawa, Shigeaki; Takeda, Takeshi; Saikusa, Akio; Furusawa, Takayuki; Takamatsu, Kuniyoshi; Nishihara, Tetsuo; Sawa, Kazuhiro; Iyoku, Tatsuo
JAERI-Tech 2002-059, 42 Pages, 2002/08
no abstracts in English
Fujimoto, Nozomu; Takada, Eiji*; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatsuo
JAERI-Tech 2001-090, 69 Pages, 2002/01
HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate showed higher results than expected value at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in a core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses.
Nakagawa, Shigeaki; Saikusa, Akio; Kunitomi, Kazuhiko
Nuclear Technology, 133(2), p.141 - 152, 2001/02
no abstracts in English
Tachibana, Yukio; Hontani, Koji*; Takeda, Takeshi; Saikusa, Akio; Shinozaki, Masayuki; Isozaki, Minoru; Iyoku, Tatsuo; Kunitomi, Kazuhiko
Nuclear Engineering and Design, 201(2-3), p.227 - 238, 2000/10
no abstracts in English