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Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Ito, Hiromichi*; Ota, Katsu; Kawahara, Hirotaka; Kobayashi, Tetsuhiko; Takamatsu, Misao; Nagai, Akinori
JAEA-Technology 2016-008, 87 Pages, 2016/05
JAEA-Technology-2016-008.pdf:18.11MB
In the experimental fast reactor Joyo, as a part of the restoration work of a partial dysfunction of fuel handling, the replacement of the upper core structure (UCS) was started from March 2014, and was completed in December 2014. In the jack-up test, the UCS was jacked-up to 1000 mm without significant sodium shearing resistance and interference. In the replacement of the UCS, a procedure was prepared with the use of wire-jack equipment assuming the interference. As a result, with the procedure and wire-jack were effectively functioned, the work was successfully completed.
Takamatsu, Misao; Kawahara, Hirotaka; Ito, Hiromichi; Ushiki, Hiroshi; Suzuki, Nobuhiro; Sasaki, Jun; Ota, Katsu; Okuda, Eiji; Kobayashi, Tetsuhiko; Nagai, Akinori; et al.
Nihon Genshiryoku Gakkai Wabun Rombunshi, 15(1), p.32 - 42, 2016/03
In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of "MARICO-2" (material testing rig with temperature control) had been broken and bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS). This paper describes the results of the in-vessel repair techniques for UCS replacement, which are developed in Joyo. UCS replacement was successfully completed in 2014. In-vessel repair techniques for sodium cooled fast reactors (SFRs) are important in confirming its safety and integrity. In order to secure the reliability of these techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. The experience and knowledge gained in UCS replacement provides valuable insights into further improvements for In-vessel repair techniques in SFRs.
Ota, Katsu; Ushiki, Hiroshi*; Maeda, Shigetaka; Kawahara, Hirotaka; Takamatsu, Misao; Kobayashi, Tetsuhiko; Kikuchi, Yuki; Tobita, Shigeharu; Nagai, Akinori
JAEA-Technology 2015-026, 180 Pages, 2015/11
JAEA-Technology-2015-026.pdf:79.87MB
In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of "MARICO-2" (material testing rig with temperature control) had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS). The replacement of the UCS was conducted from May to December 2014. The design and manufacture of UCS was started from 2008, and the installation of UCS was completed successfully in November 21th 2014. The major results gained during the design and manufacture of UCS is as follows.
Ito, Hiromichi; Suzuki, Nobuhiro; Kobayashi, Tetsuhiko; Kawahara, Hirotaka; Nagai, Akinori; Sakao, Ryuta*; Murata, Chotaro*; Tanaka, Junya*; Matsusaka, Yasunori*; Tatsuno, Takahiro*
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.1058 - 1067, 2015/05
In the experimental fast reactor Joyo (Sodium-cooled Fast Reactor (SFR)), it was confirmed that the top of the irradiation test sub-assembly had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS). There is a risk of deformation of the UCS and guide sleeve (GS) caused by interference between them unless inclination is controlled precisely. To mitigate the risk, special jack-up equipment for applying three-point suspension was developed. The existing damaged UCS (ed-UCS) jack-up test using the jack-up equipment was conducted on May 7, 2014. As a result of this test, it was confirmed that the ed-UCS could be successfully jacked-up to 1000 mm without consequent overload. The experience and knowledge gained in the ed-UCS jack-up test provides valuable insights and prospects not only for UCS replacement but also for further improving and verifying repair techniques in SFRs.
Ito, Hiromichi; Takamatsu, Misao; Kawahara, Hirotaka; Nagai, Akinori
JAEA-Technology 2014-024, 28 Pages, 2014/07
JAEA-Technology-2014-024.pdf:17.45MB
Because the gap between the UCS and rotation plug's guide sleeve is 5 mm in minimum, there is a risk of deformation of the UCS and guide sleeve with interference between UCS and guide sleeve in the UCS replacement work. In order to reduce the risk, R&D for below subjects is required.(1) UCS jack-up equipment with strict control of inclination, (2) Detection and escape method for interference between UCS and guide sleeve. In order to solve above (1), the jack-up equipment with applying three-point suspension is developed. Then, in the aspect of above (2), load-measuring devices are installed on three jacks of jack-up equipment. By means of detection eccentric load with interference, deformation of UCS and guide sleeve are prevented. And also, the location of interference can be investigated based on eccentric loads of three jacks. The performance is verified in the ex-vessel mock-up test using full-scale dummy of UCS.
Takamatsu, Misao; Ashida, Takashi; Kobayashi, Tetsuhiko; Kawahara, Hirotaka; Ito, Hideaki; Nagai, Akinori
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 10 Pages, 2013/03
In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test Sub-Assembly (S/A) of "MARICO-2" (material testing rig with temperature control) had bent onto the in-vessel storage rack (IVS) as an obstacle and had damaged the Upper Core Structure (UCS). This incident necessitates the replacement of the UCS and the retrieval of MARICO-2 S/A for Joyo re-start. Along with four stages involving jack-up and retrieval of the existing damaged UCS (ed-UCS), retrieval of the MARICO-2 S/A, and installation of the new UCS (n-UCS) in the restoration work plan, current conditions at Joyo are being carefully investigated, and the results are applied to the designs of special handling equipment, which is being manufactured and scheduled for operation in 2014.
Kawahara, Hirotaka; Yamamoto, Masaya; Tomita, Etsuo; Takamatsu, Misao
JAEA-Technology 2012-030, 50 Pages, 2012/09
JAEA-Technology-2012-030.pdf:6.16MB
JAEA-Technology-2012-030-appendix(CD-ROM).zip:21.06MB
In the experimental fast reactor Joyo, in-vessel observation results showed that 6 pins which were connected between the handling head and the wrapper tube joint of the instrumented test subassembly (MARICO-2) were disconnected. Therefore, in order to confirm whether the disconnected 6 pins will influence reactor's safety or not, loose parts behavior in the reactor vessel was evaluated.
Ito, Chikara; Isozaki, Kazunori; Ashida, Takashi; Sumino, Kozo; Kawahara, Hirotaka
IAEA-TECDOC-1633 (Internet), p.45 - 56, 2009/11
Okawachi, Yasushi; Maeda, Shigetaka; Ito, Chikara; Kawahara, Hirotaka; Aoyama, Takafumi; Ishida, Koichi
JAEA-Technology 2009-047, 130 Pages, 2009/09
JAEA-Technology-2009-047-1.pdf:44.46MBJAEA-Technology-2009-047-2.pdf:38.7MBJAEA-Technology-2009-047-3.pdf:33.16MB
This report summarizes the contents about "Reactor physics and plant dynamics experiments using the Joyo simulator" which is one of the training themes. Training is performed using the full scope nuclear reactor simulator for Joyo operation training. While pushing from starting of a nuclear reactor in each experiment of criticality, a control rod proofreading examination, measurement of the temperature of a nuclear reactor, or the reactivity coefficient accompanying output change, feedback reactivity measurement of a fast reactor, etc. and understanding self-regulating characteristics peculiar to a nuclear reactor, the operation of a nuclear reactor can be experienced.
Ishikawa, Koki; Takamatsu, Misao; Kawahara, Hirotaka; Mihara, Takatsugu; Kurisaka, Kenichi; Terano, Toshihiro; Murakami, Takanori; Noritsugi, Akihiro; Iseki, Atsushi; Saito, Takakazu; et al.
JAEA-Technology 2009-004, 140 Pages, 2009/05
JAEA-Technology-2009-004.pdf:2.0MB
Probabilistic safety assessment (PSA) has been applied to nuclear plants as a method to achieve effective safety regulation and safety management. In order to establish the PSA standard for fast breeder reactor (FBR), the FBR-PSA for internal events in rated power operation is studied by Japan Atomic Energy Agency (JAEA). The level1 PSA on the experimental fast reactor Joyo was conducted to investigate core damage probability for internal events with taking human factors effect and dependent failures into account. The result of this study shows that the core damage probability of Joyo is 5.0
10
per reactor year (/ry) and that the core damage probability is smaller than the safety goal for existed plants (10 ry) and future plants (10
/ry) in the IAEA INSAG-12 (International Nuclear Safety Advisory Group) basic safety principle.
Matsuba, Kenichi; Ito, Chikara; Kawahara, Hirotaka; Aoyama, Takafumi
Journal of Power and Energy Systems (Internet), 2(2), p.545 - 556, 2008/00
Significant thermal stresses are loaded onto the structures of sodium-cooled fast reactor (SFR) due to high temperature and large temperature gradients associated with employing sodium coolant with its high thermal conductivity and low heat capacity. Therefore, it is important to monitor the temperature variation, related stress and displacement, and vibration in the cooling system piping and components in order to assure structural integrity while the reactor plant is in-service. SFR structural integrity monitoring can be enhanced by an optical fiber sensor, which is capable of continuous or dispersed distribution measurements of various properties such as radiation dose, temperature, strain, displacement and acceleration. In the experimental fast reactor Joyo, displacement and vibration measurements of the primary cooling system have been carried out using Fiber Bragg Grating (FBG) sensors to evaluate the durability and measurement accuracy of FBG sensors in a high 
-ray environment. The data were successfully obtained with no significant signal loss up to an accumulated -ray dose of approximately 410
Gy corresponding to 120 EFPDs operation. Measured displacement of the piping support was nearly equal to the calculated thermal displacement. Measured vibration power spectra of the piping support were similar to those measured with a reference acceleration sensor. The measured results indicate that the FBG sensor is suitable for monitoring the displacement and vibration aspects of fast reactor cooling system integrity in a high -ray environment.
Matsuba, Kenichi; Ito, Chikara; Kawahara, Hirotaka; Aoyama, Takafumi
Dai-12-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.13 - 14, 2007/06
no abstracts in English
Matsuba, Kenichi; Ito, Chikara; Kawahara, Hirotaka; Aoyama, Takafumi
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
Significant thermal stresses are loaded on the structures of sodium-cooled fast reactor (SFR) due to high temperature and large temperature gradients associated with employing sodium coolant. Therefore, it is important to monitor the temperature variation and related stress on the cooling system piping in order to assure structural integrity. Structural integrity monitoring can be enhanced by an optical fiber sensor, which is capable of continuous or dispersed distribution measurements of various physical properties such as radiation dose, temperature, strain, displacement and acceleration. In the experimental fast reactor Joyo, displacement and vibration measurements of the primary cooling system have been carried out using Fiber Bragg Grating (FBG) sensors to evaluate the durability and measurement accuracy of FBG sensors in a high ray environment. The data were successfully obtained with no significant signal loss up to an accumulated ray dose of approximately 410Gy corresponding to 120EFPDs operation. Measured displacement of the piping support was nearly equal to the calculated thermal displacement. Measured vibration power spectra of the piping support were similar to those measured with a reference acceleration sensor. The measured results indicate that the FBG sensor is applicable for monitoring the displacement and vibration of fast reactor cooling system integrity in a high ray environment.
Matsuba, Kenichi; Kawahara, Hirotaka; Aoyama, Takafumi
JAEA-Conf 2006-003, p.24 - 37, 2006/05
The experimental fast reactor JOYO at O-arai Engineering Center of Japan Nuclear Cycle Development Institute is the first liquid metal cooled fast reactor in Japan. This paper describes the plant outline, experiences on the fast reactor technology and test results accumulated through twenty eight years successful operation of JOYO.
Matsuba, Kenichi; Kawahara, Hirotaka; Ito, Chikara; Yoshida, Akihiro; Nakai, Satoru
UTNL-R-0453, p.12_1 - 12_10, 2006/03
no abstracts in English
Isozaki, Kazunori; Ogawa, Toru; Nishino, Kazunari; Kaito, Yasuaki; Ichige, Satoshi; Sumino, Kozo; Suto, Masayoshi; Kawahara, Hirotaka; Suzuki, Toshiaki; Takamatsu, Misao; et al.
JNC TN9440 2005-003, 708 Pages, 2005/05
JNC-TN9440-2005-003.pdf:31.46MB
Periodic safety review (Review of the aging management) which consisted of Technical review on aging for the safety related structures, systems and components and Establish a long term maintenance program was carried out up to April 2005.1. Technical review on aging for the safety related structures, systems and components It was technically confirmed to prevent the loss of function of the safety related structures, systems and components due to aging phenomena, which (1) irradiation damage, (2) corrosion, (3) abrasion and erosion, (4) thermal aging, (5) creep and fatigue, (6) Stress Corrosion Cracking, (7) insulation deterioration and (8) general deterioration, under the periodic monitoring or renewal of them 2. Establish a long term maintenance program The long term maintenance program during JPY2005 to 2014 were established based on the technical review on aging for the safety related structures, systems and components. It was evaluated that the inspection or renewal based on the long term maintenance program, in addition to the spontaneous inspection long-term schedule of the long term voluntary inspection plan, could prevent the loss of function of the safety related structures, systems and components in future.
Ito, Keisuke; Kawahara, Hirotaka; Mori, Takero; Jo, Takahisa; Ariyoshi, Masahiko; Isozaki, Kazunori
JNC TN9410 2005-008, 267 Pages, 2005/03
JNC-TN9410-2005-008.pdf:108.4MB
Control characteristic tests of reactor coolant temperature control system were carried out to confirm its controlling constant to make MK-III hear transport system control stably, and stability against actual disturbance to the plant. The control characteristic test consists of three kinds of tests. As a result, the optimum PI parameters of the reactor coolant temperature control system was confirmed that the proportional gain is between from 0.36 to 1.12(approximately half of MK-II), the integral time is 80 respectively. The gain margin of the control system was between from 7 to 19dB of through the vane opening range.
Oyama, Kazuhiro; Kawahara, Hirotaka; Ishida, Koichi; Ariyoshi, Masahiko; Isozaki, Kazunori; Sugaya, Kazushi*; Fukami, Akihiro*
JNC TN9410 2005-006, 121 Pages, 2005/03
JNC-TN9410-2005-006.pdf:10.81MB
In the MK-III performance test, the experimental fast reactor JOYO raised the reactor thermal power gradually with about 20%, 50%, 75%, 90%, and 100% (140MWt), and reached 140MWt which are the full power of a MK-III reactor core on October 28, 2003. Then, continuation operation beyond full power 100 hour was attained. This report summarized the result of power-up test , full power continuation operation test, blower start-up test.The outline is as follows.(1)From the standby state (system temperature of 250degree), the usual power-up operation (an power-up rate ;about 5MWt/20min, a power is held for about 10 minutes every 5MWt) attained the reactor thermal full power (140MWt) gradually on October 28,2004. Moreover, it checked that each part temperature and flow were less than alarm setting values on each power level.(2)The reactor thermal power was made into the parameter, a series of operations about the blower start-up, and the influence which it has on coolant temperature was checked. As a result, the optimal reactor thermal power which starts up the blower from a natural ventilation cooling state was set to about 18 MWt, and the starting procedure was made into the method(order of 1A-2A-1B-2B) which starts four sets of the one blower at a time one by one.(3)It checked that reactor shutdown operation by two control-rod simultaneous insertion at 35MWt, and it could carry out with time margin with a series of sufficient operations of resulting from control rod insertion in the blower shutdown. By adopting this reactor shutdown operation method, operation of an operation stuff was mitigated and it checked that plant characteristics also improved.(4)The reactor full power was reached on November 14. Continuation operation beyond full power 100 hour was attained after that till on November 20, 10:30. The data of each part of a plant was acquired at intervals of 24 hours, and it checked that it was less than an alarm setting value.
Oyama, Kazuhiro; Kawahara, Hirotaka; Ariyoshi, Masahiko; Isozaki, Kazunori; Sugaya, Kazushi*; Fukami, Akihiro*
JNC TN9410 2005-005, 56 Pages, 2005/03
JNC-TN9410-2005-005.pdf:14.56MB
The experimental fast reactor JOYO MK-III increased that the reactor thermal power by the factor 1.4. The main intermediate heat exchangers (IHX) and the dump heat exchangers (DHX) were exchanged. And then, the flow rate of the main cooling system, the secondary cooling system were increased. As one of the performance test to confirm that the cooling system which included these switch receptacles has an enough decay heat performance, it did an heat transfer characteristics test and it evaluated a heat balance, the decay heat performance of IHX and DHX.The outline is as follows.(1)It confirmed that the modificated plant had fixed performance by the heat balance of full power.(2)The secondary inlet temperature of B-loop IHX is about 6degree higher out of the cooling system with A-loop. It thinks that this is one because of the difference ( about 2 % ) with the flow rate of the main cooling system in measurement. There was decay heat capacity of the A-loop and the B-loop in the balance, making the flow rate of the main cooling system of the A-loop positive and supposing that the B-loop is a revision flow rate and as for the heat transfer performance of IHX of the A-loop and the B-loop, the approximately equal thing could be confirmed. As a result, as for the overall heat transfer coefficient of IHX, the A-loop was about 125 % of the design value, the B-loop was about 129 % of the design value and it confirmed that two IHX had the performance to be equal and an enough decay heat performance.(3)It made DHX outlet air temperature about 20degree and it calculated DHX outlet air flow from the decay heat capacity from sodium coolant and the DHX outlet air temperature in full power. As a result, DHX could confirm that the decay heat ability to be equivalent to he reactor thermal power in 85 - 90 % of capacities of design value (6,750m
/min) and an enough decay heat performance.
