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Yada, Hiroki; Takaya, Shigeru; Wakai, Takashi; Nakai, Satoru; Machida, Hideo*
Nihon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00389_1 - 17-00389_15, 2018/03
no abstracts in English
Nakai, Satoru
Dekomisshoningu Giho, (56), p.14 - 28, 2017/09
Prototype fast breeder reactor power plant Monju which is under construction was decided by the Japanese government not to operate but to be decommissioned safely and surely in December 2016. In the view point of decommissioning, one of the major difference from LWR is sodium as a coolant. In the overseas such as U.K., Germany, the United States, France, there is the precedent example of decommissioning and can be referred to it. In this report, the situation and problem of overseas example about removal and disposal of sodium.
Nakai, Satoru
Fast Reactor System Design, p.249 - 267, 2017/03
The atomic energy plant has to maintain safety, reliability and structural integrity through plant life. Therefore, careful operation such as avoiding the thermal stress deviated from a design condition caused by a rapid temperature change is necessary. In addition, by the huge complexity system such as the nuclear power plant, a prediction of behavior during the life at the design stage is accompanied with uncertainty, and it is difficult to secure safety, reliability for a plant life only by a design. Therefore, appropriate maintenance activity is necessary, and consideration to the maintenance in the design stage relatively grows important. Particularly, the importance becomes still larger because uncertainty is big about the new type reactor. Therefore, I think that I want you to learn a way of thinking about the maintenance that is based on the characteristic of the fast reactor and basics of the maintenance of the nuclear power plant which is a huge complexity system.
Takaya, Shigeru; Asayama, Tai; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Nakai, Satoru; Morishita, Masaki
Journal of Nuclear Engineering and Radiation Science, 1(1), p.011004_1 - 011004_9, 2015/01
A new process for determination of inservice inspection (ISI) requirements was proposed based on the System Based Code concept to realize effective and rational ISI by properly taking into account plant specific features. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other one on detectability of defects before they would grow to an unacceptable size in light of plant safety. If defect detection was not feasible, structural integrity evaluation would be required under sufficiently conservative hypothesis. The applicability of the proposed process was illustrated through an application to the existing prototype fast breeder reactor, Monju.
Nakai, Satoru
Genshiryoku Kyokasho "Kosokuro Shisutemu Sekkei", p.199 - 214, 2014/09
The atomic energy plant has to maintain safety, reliability and structural integrity through plant life. Therefore, careful operation such as avoiding the thermal stress deviated from a design condition caused by a rapid temperature change is necessary. In addition, by the huge complexity system such as the nuclear power plant, a prediction of behavior during the life at the design stage is accompanied with uncertainty, and it is difficult to secure safety, reliability for a plant life only by a design. Therefore, appropriate maintenance activity is necessary, and consideration to the maintenance in the design stage relatively grows important. Particularly, the importance becomes still larger because uncertainty is big about the new type reactor. Therefore, I think that I want you to learn a way of thinking about the maintenance that is based on the characteristic of the fast reactor and basics of the maintenance of the nuclear power plant which is a huge complexity system.
Nakai, Satoru
Hozengaku, 13(2), p.41 - 42, 2014/07
The road map to establish a fast reactor (FR) maintenance technology in the technical aspect became clear in the examination of the FR maintenance in the Japan Society of Maintenology (JSM). It is vital to acquire operation and maintenance experience of the plant to implement this road map, and to establish a fast reactor maintenance technology. On the other hand, even if components to be maintained are defined and the maintenance methods are established, improvement of the maintenance ability in the organizations and individuals, proper and reliable implementation of the maintenance are indispensable. To improve the ability for maintenance, the action of the organized education, training and technical transmission is necessary including a light water reactor. The examination of the FR maintenance technology is a good opportunity for the application of the maintenance principles established by the JSM to the FR.
Nakai, Satoru; Kaneko, Yoshihisa; Mukai, Kazuo
Hozengaku, 9(4), p.44 - 49, 2011/01
The sodium leak accident in a secondary main cooling system of prototype fast breeder reactor Monju on December 8th in 1995, and the Monju has shut down for 14 and half years since that. JAEA improved the safety by investigating the sodium leakage accident, taking countermeasures such as better understanding by the local's, improvement of operation. In the maintenance field, the confirmation of Monju systems and components integrity, systems improvement as sodium leak countermeasures, introduction of the maintenance program to resolve maintenance problems. Monju restarted on May 6th 2010 after the confirmation of restart readiness by safety authorities and the core confirmation test which is first test after restart was continued to July 22nd as planned.
Sakanaka, Shogo*; Akemoto, Mitsuo*; Aoto, Tomohiro*; Arakawa, Dai*; Asaoka, Seiji*; Enomoto, Atsushi*; Fukuda, Shigeki*; Furukawa, Kazuro*; Furuya, Takaaki*; Haga, Kaiichi*; et al.
Proceedings of 1st International Particle Accelerator Conference (IPAC '10) (Internet), p.2338 - 2340, 2010/05
Future synchrotron light source using a 5-GeV energy recovery linac (ERL) is under proposal by our Japanese collaboration team, and we are conducting R&D efforts for that. We are developing high-brightness DC photocathode guns, two types of cryomodules for both injector and main superconducting (SC) linacs, and 1.3 GHz high CW-power RF sources. We are also constructing the Compact ERL (cERL) for demonstrating the recirculation of low-emittance, high-current beams using above-mentioned critical technologies.
Nakai, Satoru; Aoyama, Takafumi; Ito, Chikara; Yamamoto, Masaya; Iijima, Minoru; Nagaoki, Yoshihiro; Kobayashi, Atsuko; Onoda, Yuichi; Ohgama, Kazuya; Uwaba, Tomoyuki; et al.
Kosoku Jikkenro "Joyo" Rinkai 30-Shunen Kinen Hokokukai Oyobi Gijutsu Koenkai, 154 Pages, 2008/06
no abstracts in English
Aoyama, Takafumi; Sekine, Takashi; Nakai, Satoru; Suzuki, Soju
Proceedings of 15th Pacific Basin Nuclear Conference (PBNC-15) (CD-ROM), 6 Pages, 2006/10
The experimental fast reactor JOYO is the first liquid sodium fast reactor in Japan. The purpose of constructing JOYO was to obtain technical information about liquid metal fast breeder reactors (LMFBR). In addition to providing operating experience, many kinds of irradiation tests have been conducted for the development of fuels and materials under the conditions of higher fast neutron flux and temperature than those in LWRs. JOYO has been operated successfully since its criticality was first achieved in 1977 without any serious problem, and this operation demonstrated the safety and reliability of the sodium cooled fast reactor. Continual facility improvements have been punctuated by major enhancements, the latest of which is MK-III. Compared to MK-II, MK-III has a four times larger irradiation capability, improved irradiation test vehicles and improved irradiation characterization. The applications of this enhanced capability include testing fuels and safety features for future FBRs, exploring use of fast reactors for transmutation of radioactive waste, and developing advanced materials for fusion power. In light of the shutdown of several fast reactors around the world, the ability to make such major contributions to reactor development takes on even greater significance. Irradiation tests, steady-state and safety related operations of JOYO are also expected to promote the development of JAEA's prototype FBR, Monju.
Matsuba, Kenichi; Kawahara, Hirotaka; Ito, Chikara; Yoshida, Akihiro; Nakai, Satoru
UTNL-R-0453, p.12_1 - 12_10, 2006/03
no abstracts in English
Sumino, Kozo; Isozaki, Kazunori; Ashida, Takashi;
Nuclear Technology, 150(1), p.56 - 66, 2005/04
Times Cited Count：5 Percentile：30.51(Nuclear Science & Technology)Focusing on the cover layer materials (as the Radon Barrier Materials), which could have the effect to restrain the radon from scattering into the air and the effect of the radiation shielding, we produced the radon barrier materials with crude bentonite on an experimental basis, using the rotary type comprehensive unit for grinding and mixing, through which we carried out the evaluation of the characteristics thereof.
Sumino, Kozo; Ashida, Takashi; Kawahara, Hirotaka; Ichige, Satoshi; Isozaki, Kazunori; Nakai, Satoru
Proceedings of Operating Nuclear Facility Safety(2004ONFS),p204-216, p.204 - 216, 2004/11
None
Kawahara, Hirotaka; Isozaki, Kazunori; Ishii, Takayuki; Ichige, Satoshi; Nose, Shoichi; Sakaba, Hideo; Nakai, Satoru
JNC TN9410 2004-016, 106 Pages, 2004/06
JNC-TN9410-2004-016.pdf:8.47MB
A key part of the upgrade of the experimental fast reactor JOYO to the MK-III design was the replacement of the dump heat exchangers. MK-III function tests (SKS-1) of the new dump heat exchangers were carried out from August 27,2001 through September 13,2001. The major results of the function tests of the dump heat exchangers were as follows: (1) Air flow of the main blower with an inlet vane opening of 50% was confirmed to exceed the design rated flow of 7,700m3/min. It was also demonstrated that an inlet vane opening of 100% provides about 130% of the design rated flow. This is because the new DHX flow route has more low pressure loss than the design value. (2) Tests of the air flow of the main blower demonstrated that with a fully opened inlet damper a full opened outlet damper and an inlet vane opening of O% provides about 5% of the design rated flow. (3) Free flow coast down characteristics of the main blower achieved an inlet vane O% opening in an average of 7.9 seconds. Revolutions per minute of the main blower reached zero in an average of 8.7 seconds. The delay time from the opening of the vacuum contact breaker to the air flow decrease was approximately 1 second. This was a more conservative value than the 5 seconds assumed in design thermal transient analyses. (4) The loudest noise occurred with the main blower operating with a 25% inlet vane opening. At that time, the noise around the main blower was approximately 100dB, and in the surrounding monitoring area boundary, the noise was 50dB. This was confirmed to be within the standard of the Ibaraki prefectural ordinance. (5) Although the MK-III inlet vane and inlet damper drive unit was bigger than the MK-II unit, the accumulator tank was confirmed to provide sufficient volume during a compression air loss event.
Kawahara, Hirotaka; Kawahara, Hirotaka; Ichige, Satoshi; Isozaki, Kazunori; Nakai, Satoru
Proceedings of 12th International Conference on Nuclear Engineering (ICONE-12) (CD-ROM), 0 Pages, 2004/00
A recently completed major upgrade of the JOYO experimental sodium-cooled fast reactor, to the MK-III design, increased its irradiation capability approximately four times. 0ne major change was a 40% increase in thermal power to 140 MWt, which necessitated the replacement of the heat exchangers. Each of the two coolant loops includes an intermediate heat exchanger (IHX) and sodium pump in the primary system, and two dump heat exchangers (DHXs) and a pump in the secondary system. The heat transfer area of the finned tubes in each (air-cooled) DHX was doubled, compared to the old design, to achieve a 35 MWt rating, Major challenges in the replacement of secondary components, such as piping and DHX, were control of impurity ingress into the sodium system, and integrity assurance of the welding. Damage to existing components and systems was avoided during cutting and welding operations by taking measures to Prevent ingress of air into the sodium systems. The measures included use of seal b
; Kawasaki, Hirotsugu; *;
JNC TN9410 2002-004, 100 Pages, 2002/03
JNC-TN9410-2002-004.pdf:3.93MB
It prepares for a large amount of radioactive sodium processing accompanied by final shutdown/decommissioning of First Reactor Plant, and/or dismantling of sodium experimental facilities in a domestic one, and the technical development for sodium processing safely, efficiently, and economically is carried out in the future. The sodium processing method with caustic soda, which has application in the actual sodium processing of abroad, was adopted, and the experimental research was started aiming at establishment of the fundamental processig method when applying this in our country. Then, the sodium conversion test apparatus "SCOT" aiming at grasp of the basic reaction property of sodium and caustic soda, and the optimum specification determination of a sodium conversion processing system, was designed and fabricated. As a result of checking the system function of this apparatus and carrying out a small reaction test as a previously test run, the following findings and subjects of future were obtained. (1)In the functional test of each performance in this apparatus, the design performance was sufficiently satisfied on each function and controllability. (2)In the calibration test of a caustic soda concentration meter, the correlation data between ultrasonic velocity and caustic soda concentration as a function of caustic soda temperature was obtained in the condition range extended conventionally, and the calibration curve was proposed by multiple regression analysis. (3)In the reaction test which carries out very-small-quantity pouring of the sodium at caustic soda, the problem on the safety was not recognized a reaction phenomenon or a process behavior. However, some problems, such as fluctuation of the sodium flow meter indication by noise influence, plugging in the spray nozzle, and involvement atomizing gas in a caustic soda circulation line, were occurred in this test run. In the future, it will be coped with from both sides of the equipment modificatkm ...
Maeda, Yukimoto; Aoyama, Takafumi; Odo, Toshihiro; Nakai, Satoru;
P96, 96 Pages, 2002/00
None
Hirakawa, Yasushi; *; ;
JNC TN9400 2001-079, 43 Pages, 2001/03
JNC-TN9400-2001-079.pdf:2.06MB
To establish sodium removal technology is essential from the viewpoints of low radiation of works and low environmental effect at the time of maintenance, repair and decommissioning of LMFBRs. In this study, as a start of sodium removal basic test, we carried out the experiment by sodium removal basic reaction test apparatus. Molten sodium potassium alloy (NaK) as a simulated molten sodium and solid sodium were used to grasp the reaction speed of gas containing moisture and NaK or sodium. Followings are the major results: (1)Under low temperature and humidity nitrogen gas condition, molten NaK reaction rate was about twice as high as solid sodium. This result suggests the possibility of effect of sodium phase (molten or solid) to reaction rate. (2)Under relatively high temperature and humidity nitrogen gas condition, the reaction rate of molten NaK and moist nitrogen gas was about 3 times as high as low temperature and humidity nitrogen gas condition. Gas condition affects the reaction rate largely. (3)NaK and moist carbon dioxide gas could react only the surface of the NaK and the reaction could not progress to the inside of NaK. This result shows the bulk NaK could not be converted to NaK-carbonate.
; ; ; Yatabe, Toshio
JNC TN9450 2001-004, 136 Pages, 2001/01
JNC-TN9450-2001-004.pdf:3.28MB
DNB (Departure from Nucleate Boiling) test were executed by a 1MWt Double-Wall-Tube Steam Generator. This data report describes the temperature fluctuation of the outer tube and sodium around DNB region. Furthermore, this report includes the temperature fluctuation of the inner surface of the inner tube obtained by removing noise of the original DNB signal and calculating heat flux of the tube, too. It also mentions the influence of the test parameter such as water flow rate on DNB period and DNB region length. All the DNB data described in this report were recorded by the data acquisition system of the small steam generator test facility. The contents are as followings : (1)1MW double-wall-tube steam generator, test method and test condition (2)The length of DNB region (3)DNB temperature fluctuation of the outer tube and sodium (experimental data) (4)The spectrum of DNB temperature fluctuation (5)DNB temperature fluctuation of the inner surface of the inner tube (calculation data)
