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Journal Articles

Proposal on LBB evaluation conditions for sodium cooled fast reactor pipes and effects of pipe parameters

Yada, Hiroki; Takaya, Shigeru; Wakai, Takashi; Nakai, Satoru; Machida, Hideo*

Nippon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00389_1 - 17-00389_15, 2018/03

no abstracts in English

Journal Articles

Current status and issues of sodium removal and disposal from LMFR in the framework of decommissioning

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.

Journal Articles

Maintenance

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.

Journal Articles

Application of the system based code concept to the determination of in-service inspection requirements

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.

Journal Articles

Maintenance

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.

Journal Articles

Application of maintenology to the fast reactor

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.

Journal Articles

Restart of protype FBR "Monju" after long-term shut-down

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.

Journal Articles

Celebration of 30th anniversary of the experimental fast reactor Joyo

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

Journal Articles

The H-Invitational Database (H-InvDB); A Comprehensive annotation resource for human genes and transcripts

Yamasaki, Chisato*; Murakami, Katsuhiko*; Fujii, Yasuyuki*; Sato, Yoshiharu*; Harada, Erimi*; Takeda, Junichi*; Taniya, Takayuki*; Sakate, Ryuichi*; Kikugawa, Shingo*; Shimada, Makoto*; et al.

Nucleic Acids Research, 36(Database), p.D793 - D799, 2008/01

 Times Cited Count:50 Percentile:24.18(Biochemistry & Molecular Biology)

Here we report the new features and improvements in our latest release of the H-Invitational Database, a comprehensive annotation resource for human genes and transcripts. H-InvDB, originally developed as an integrated database of the human transcriptome based on extensive annotation of large sets of fulllength cDNA (FLcDNA) clones, now provides annotation for 120 558 human mRNAs extracted from the International Nucleotide Sequence Databases (INSD), in addition to 54 978 human FLcDNAs, in the latest release H-InvDB. We mapped those human transcripts onto the human genome sequences (NCBI build 36.1) and determined 34 699 human gene clusters, which could define 34 057 protein-coding and 642 non-protein-coding loci; 858 transcribed loci overlapped with predicted pseudogenes.

Journal Articles

JOYO, the irradiation and demonstration test facility of FBR development

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.

Journal Articles

Development of a plant structure integrity monitoring system for a fast reactor based on optical fiber technology

Matsuba, Kenichi; Kawahara, Hirotaka; Ito, Chikara; Yoshida, Akihiro; Nakai, Satoru

UTNL-R-0453, p.12_1 - 12_10, 2006/03

no abstracts in English

Journal Articles

Design and renovation of heat transport system in the experimental fast reactor JOYO

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

JAEA Reports

MK-III Function Tests in JOYO; Dump Heat Exchanger (DHX)

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.

Journal Articles

Replacement of Secondary Heat Transport System Components In the Experimental Fast Reactor JOYO

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

Journal Articles

Glorious achievement of a quarter century operation and a promising project named MK-III in JOYO

Maeda, Yukimoto; Aoyama, Takafumi; ; Nakai, Satoru;

P96, 96 Pages, 2002/00

None

JAEA Reports

Large-scaled thermohydraulic tests plan for cooling systems in fast reactors; Experimental models of reactor vessel and the primary cooling system

Kamide, Hideki; Hayashi, Kenji; Gunji, Minoru; Hayashida, Hitoshi; Nishimura, Motohiko; Iitsuka, Toru; Kimura, Nobuyuki; Tanaka, Masaaki; Nakai, Satoru; Mochizuki, Hiroyasu; et al.

PNC-TN9410 96-279, 51 Pages, 1996/08

PNC-TN9410-96-279.pdf:2.92MB

Large-scaled thermohydraulic tests are planned for some new key technologies in the heat transport systems of demonstration fast reactors, in which the reactor vessel, the primary system, the secondary system, water-steam system, and the decay heat removal systems are modeled. Thermohydraulic issues and structural integrity issues were discussed for the top entry piping systems with satellite pools of the intermediate heat exchangers and the pumps, the natural circulation decay heat removal using direct heat exchangers in a reactor hot pool, the reactor vessel wall cooling system, and the new type of steam generators in the demonstration reactor. Concepts of the experimental model for the reactor vessel and the primary system were created and compared with each other for the sodium test facility which enables to answer the thermohydraulic and structural integrity issues. Following items were considered in the creation and in the selection of the models; (1)solution of the issues for Demonstration First Reactor on total system characteristics, the reactor vessel wall cooling system, the decay heat removal system, and the steam generator, (2)balance between the thermohydraulic issues and the structural integrity issues, (3)simulations of compound phenomena and interactions between the components and the heat transport systems. Total system of test facility was specified based on the selected test model.

JAEA Reports

Development of plant dynamics analysis code Super-COPD; Validation by MONJU function test

Ohtaki, Akira; Miyakawa, Akira; Nakai, Satoru

PNC-TN9410 95-060, 204 Pages, 1995/02

PNC-TN9410-95-060.pdf:7.96MB

The modular integrated plant dynamics analysis code "Super-COPD" is being developed. The validation of the code by using the MONJU functional test data was carried out to evaluate the accuracy of the calculation models and the validation results that had been carried out by using scale model test data. The procedure of the validation was as follows: (1)Setting up of the MONJU system data. The plant system data were set up by the tests data. (2)Refinement of the I/O. An input/output program for the MONJU test data was added to the code. (3)Evaluation of the test data. The test data to be used for the validation and program module to be validated were identified. The intermediate heat exchanger (IHX), steam generator (SG) and auxiliary cooling system (ACS) were selected as validation components. (4)Validation. The performance of some components was evaluated. The calculation and modification of the code were carried out. The following results were obtained. (1)IHX. The recommended calculation mesh number and plenum heat capacity and the items to be validated by the start up test were obtained. (2)SG. The recommended heat loss, validity of plenum model and the effectiveness of the flow rate to the thermal calculation were confirmed. A new model for the change of effective flow area in the SG may be needed at very low flow rate. (3)ACS. The ACS thermal/hydraulics were evaluated. The accuracy of air cooler thermal/hydraulics models, interlock/control system models, vane/blower models were validated. The accuracy of the calculation models and the validation results by using scale models was confirmed. Validation of the code using the start up test results under power operation is scheduled.

JAEA Reports

Development of Double-Wall-Tube stream generator, 10; Evaluation of thermal hydraulic tests

Nakai, Satoru; Sato, Hiroyuki

PNC-TN9410 93-294, 126 Pages, 1993/11

PNC-TN9410-93-294.pdf:4.3MB

The 1MW test model of double wall tube steam generator (DWTSG) was installed and experiments have been being carried out to evaluate the feasibility of DWTSG. Main objectives of this test are to evaluate DWTSG thermal hydraulic performance and to obtain basic data to be reflected to structural integrity of the double wall tube under steam generator circumstances such as DNB and flow instability and capability of leak detection. The evaluation of thermal hydraulic tests such as watcr single phase flow tests, two phase flow tests and boiling/super heated steam tests have been conducted. An accuracy of one dimensional thermal hydraulics analysis code, POPAI-6, has also been evaluated by the experimental data. Following results have been obtained. (1)As for the sodium side heat transfer coefficient, some modification of Graber-Rieger's correlation that was used at the DWTSG design is needed to adjust experimental data. Modification factor is 0.8. (2)Roko's and Kon'kov's DNB quality correlations can predict experimental results in $$pm$$15% error. (3)The modification factor of Dittus-Boelter is 0.8 in pre-heat region. (4)The experimental heat transfer coefficicnt has a wide range and it is very difficult to derive an experimental correlation. (5)The modification factors of Bishop and Mod. Tong are 0.65 and 0.8 respectively in post dryout region. (6)The modification factors of Bishop and Dittus-Boelter are 0.73 and 0.58, respectively, in steam region. (7)At DNB point, the Jens-Lottes correlation gives conservativc heat transfer coefficient as a nucleate boiling and the Bishop correlation with abovc modification factor gives conservative heat transfer coefficient as post dry out heat transfer coefficient. (8)Experimental heat transfer correlation were derived besides nucleate boiling. (9)Both sodium and water/steam side temperature distribution is in a predicted at the normal operation. However the less or sodium/water flow rate ratio, the larger steam temperature ...

Oral presentation

Periodical safety review and management of aged research reactor

Nakai, Satoru

no journal, , 

no abstracts in English

Oral presentation

Confirmation of Monju systems and components integrity and maintenance program toward Monju restart

Nakai, Satoru; Mukai, Kazuo

no journal, , 

no abstracts in English

28 (Records 1-20 displayed on this page)