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JAEA Reports

Assessment report of research and development activities in FY2014; Activity "Fusion Research and Development" (In-advance evaluation)

Fusion Research and Development Directorate

JAEA-Evaluation 2016-002, 40 Pages, 2016/03

JAEA-Evaluation-2016-002.pdf:2.66MB

Japan Atomic Energy Agency (hereinafter referred to as "JAEA") asked the assessment committee, "Evaluation Committee of Research and Development Activities for Fusion" (hereinafter referred to as "Committee") for in-advance evaluation of "Research and Development of the technical system for extraction of fusion energy," in accordance with "General Guideline for the Evaluation of Government R&D Activities" by Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and "Regulation on Conduct for Evaluation of R&D Activities" by JAEA. In response to the JAEA's request, the Committee assessed the research program of the Fusion Research and Development Directorate (hereinafter referred to as "FRDD") during the period from April 2015 to March 2022. The Committee evaluated the management and research activities of the FRDD based on the explanatory documents prepared by the FRDD, the oral presentations with questions-and-answers by the Director General and the Deputy Director Generals.

JAEA Reports

Assessment report of research and development activities in FY2014; Activity "Fusion Research and Development" (Result evaluation)

Fusion Research and Development Directorate

JAEA-Evaluation 2016-001, 128 Pages, 2016/03

JAEA-Evaluation-2016-001.pdf:33.25MB

Japan Atomic Energy Agency (hereinafter referred to as "JAEA") asked the assessment committee, "Evaluation Committee of Research and Development Activities for Fusion" (hereinafter referred to as "Committee") for result evaluation of "Research and Development of the Technical System for Extraction of Fusion Energy," in accordance with "General Guideline for the Evaluation of Government R&D Activities" by Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology " and "Regulation on Conduct for Evaluation of R&D Activities" by JAEA. In response to the JAEA's request, the Committee assessed the research program of the Fusion Research and Development Directorate (hereinafter referred to as "FRDD") during the period from April 2010 to November 2014. The Committee evaluated the management and research activities of the FRDD based on the explanatory documents prepared by the FRDD, the oral presentations with questions-and-answers by the Director General and the Deputy Director Generals.

Journal Articles

Status and prospect of JT-60 plasma control and diagnostic data processing systems for advanced operation scenarios

Kurihara, Kenichi; Yonekawa, Izuru; Kawamata, Yoichi; Sueoka, Michiharu; Hosoyama, Hiroki*; Sakata, Shinya; Oshima, Takayuki; Sato, Minoru; Kiyono, Kimihiro; Ozeki, Takahisa

Fusion Engineering and Design, 81(15-17), p.1729 - 1734, 2006/07

 Times Cited Count:13 Percentile:30.08(Nuclear Science & Technology)

A large tokamak fusion device JT-60 is expected to explore more advanced tokamak discharge scenario towards the ITER and a future power reactor. We believe the following experimental issues are expected to be solved in JT-60. To clarify how to keep a steady-state plasma with high performance, and how to avoid plasma instabilities almost completely. By stimulus of this motivation, several essential development and modifications of plasma control and data acquisition systems have been performed in JT-60. In this report, we discuss the developments to improve the JT-60 plasma control and data acquisition systems. In addition, a future plasma control and data acquisition systems leading to a standard design for a power reactor is envisaged on the basis of the 20-year plasma operation experiences.

Journal Articles

Status and prospect of JT-60 plasma control system for advanced tokamak discharge scenarios

Kurihara, Kenichi; JT-60 Team

Proceedings of 9th International Conference on Accelerator and Large Experimental Physics Control Systems (ICALEPCS 2003) (CD-ROM), p.612 - 616, 2004/10

Since tokamak magnetic fusion research has just made a step forward to an international collaborative project ITER, the existing tokamaks including JT-60 are expected to explore more advanced operation scenarios toward the ITER and a future power reactor. Hence, we believe the following experimental issues are to be adequately discussed, and possibly to be solved in JT-60: To clarify how to keep a steady-state plasma with high performance, and how to avoid plasma instabilities almost completely. By stimulus of this motivation, the JT-60 plasma real-time control system has been drastically improved with remodeling in hardware as well as in software. In particular, several essential developments for exploring advanced scenarios have been accomplished or are being conducted. In the developments, we adopted the fast board computers' network linking through large reflective memories. Finally, in addition to presentation on plasma control engineering activities in JT-60, we would like to envisage a future plasma control system toward a power reactor.

Journal Articles

Outline of the ITER project

Mori, Masahiro

Koon Gakkai-Shi, 30(5), p.236 - 242, 2004/09

no abstracts in English

Journal Articles

ITER engineering design

Shimomura, Yasuo; Tsunematsu, Toshihide; Yamamoto, Shin; Maruyama, So; Mizoguchi, Tadanori*; Takahashi, Yoshikazu; Yoshida, Kiyoshi; Kitamura, Kazunori*; Ioki, Kimihiro*; Inoue, Takashi; et al.

Purazuma, Kaku Yugo Gakkai-Shi, 78(Suppl.), 224 Pages, 2002/01

no abstracts in English

Journal Articles

Fusion reactor technology; Challenge to future energy

Seki, Masahiro; Hishinuma, Akimichi; Kurihara, Kenichi; Akiba, Masato; Abe, Tetsuya; Ishitsuka, Etsuo; Imai, Tsuyoshi; Enoeda, Mikio; Ohira, Shigeru; Okumura, Yoshikazu; et al.

Kaku Yugoro Kogaku Gairon; Mirai Enerugi Eno Chosen, 246 Pages, 2001/09

no abstracts in English

JAEA Reports

Validation of sodium fire analysis code ASSCOPS

Ohno, Shuji; Matsuki, Takuo*

JNC-TN9400 2000-106, 132 Pages, 2000/12

JNC-TN9400-2000-106.pdf:2.8MB

Sodium fire analyses were performed on 7 kinds of sodium leak tests using a computer code ASSCOPS which has been developed to evaluate thermal consequences of sodium leak accident in an FBR plant. By the comparison between the calculated and the test results of gas pressure, gas temperature, sodium catch pan temperature, wall temperature, and of oxygen concentration, it was confirmed that the ASSCOPS code and the parameters used in the analysis give valid or conservative results on thermal consequences of sodium leak and fire.

Journal Articles

Crack propagation tests of HIPed DSCu/SS joints for plasma facing components

Hatano, Toshihisa; Goto, Masahiro*; Yamada, Tetsuji*; Nomura, Yuichiro*; Saito, Masakatsu*

Fusion Engineering and Design, 49-50, p.207 - 212, 2000/11

 Times Cited Count:3 Percentile:72.05(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Design and analysis of the vacuum vessel for RTO/RC-ITER

Onozuka, Masanori*; Ioki, Kimihiro*; Johnson, G.*; Kodama, T.*; Sonnazzaro, G.*; Utin, Y.*

Fusion Engineering and Design, 51-52(Part.B), p.249 - 255, 2000/11

 Times Cited Count:5 Percentile:60.49(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Post irradiation examination of (U,Pu) C and (U,Pu) N fuel for fast reactor; Non-destructive examination result of the fuel pin

; ; ; Matsumoto, Shinichiro

JNC-TN9410 2000-009, 65 Pages, 2000/09

JNC-TN9410-2000-009.pdf:4.36MB

In order to evaluate irradiation behavior of(U, Pu) C and (U, Pu) N fuel using fast reactor, (U, Pu) C and (U, Pu) N fuel pins were irradiated in JOYO for the fist time in Japan. In this study, one (U, Pu) C fuel pin and two (U, Pu) N fuel pins were irradiated to maximum burn up about 40GWd/t. Post irradiation examination of (U, Pu) C and (U, Pu) N fuel pins started in Fuel Monitoring Facility (FMF) at JNC from October 1999, and it ended in March, 2000. The results of non-destructive post irradiation examination reported in this document. Main results are shown in the following. (1)The soundness of all (U,Pu) C and (U,Pu) N fuel pins were confirmed from the non-destructive examination result. (2)The fuel stack elongation of (U,Pu) C and (U,Pu) N is bigger than it of the MOX fuel for fast reactor. (3)The singular behavior from the gamma ray scanning measurement in the stack area was not confirmed. The migration of Cs137 to lower insulator pellet and outside of the pellet was confirmed in (U,Pu) N B9NO2 pin. In (U,Pu) C fuel, the migration of Cs137 was not confirmed. (4)In (U,Pu) C B9CO1 pin and (U,Pu) N B9NO2 pin in which the gap width was small, diameter of cladding increase around 50 $$mu$$m in the stack area which originates for FCMI was confirmed. In (U,Pu) N B9NO1 pin in which the gap width was wide, the ovality which originates from the relocation of the pellet was confirmed. (5)Fission gas release rate of (U,Pu) N were 3.3% and 5.2%, and the low value compared to the MOX fuel was shown.

JAEA Reports

Conceptual design of neutron shield for ECH launcher on D-T fusion reactors

Takahashi, Koji; Imai, Tsuyoshi; Mori, Kensuke*; Mori, Seiji*; Nomoto, Yasunobu*

JAERI-Research 2000-036, 26 Pages, 2000/09

JAERI-Research-2000-036.pdf:1.24MB

no abstracts in English

JAEA Reports

Irradiation tests report of the 35th cycle in "JOYO"

*

JNC-TN9440 2000-008, 79 Pages, 2000/08

JNC-TN9440-2000-008.pdf:2.33MB

This report summarizes the operating and irradiatlon data of the experimental reactor "JOYO" 35th cycle. Irradiation tests in the 35th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (2)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (3)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (4)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large scale reactor (5)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (6)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confimation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (based on a contract with universities) The maximum burnup driver assembly "PFD253" reached 67,600 MWd/t (pin average).

JAEA Reports

Evaluation for the transient Burst property of austenitic steel fuel Claddings irradiated as the MONJU type Fuel Assemblies (MFA-1&MFA-2)in FFTF

; ; Sakamoto, Naoki; *; Akasaka, Naoaki;

JNC-TN9400 2000-095, 110 Pages, 2000/07

JNC-TN9400-2000-095.pdf:13.57MB

The effects of high fluence irradiation and swelling on the transient burst properties of austenitic steel fuel claddings; PNC316 and 15Cr-20Ni stcel, which were irradiated as the MONJU type fuel assemblies (MFA-1&MFA-2) in the FFTF reactor, were investigated. The temperature-transient-to-burst tests were conducted on a total of eight irradiation conditions. Fractographic examination and TEM observation were performed in order to evaluate the effect of high dose irradiation on the transient burst property and the relation between failure mechanism and microstructural change during rapid (ramp) heating. The results of the PIE showed that there was no significant effect of irradiation on the transient burst properties of these fuel claddings under the irradiation conditions examined. the results obtained in this study are as follows; (1)The rupture temperature of the irradiated PNC316 fuel cladding of MFA-1 was as same as that of our previous works for the fluence range up to 2.13$$times$$10$$^{27}$$ n/m$$^{2}$$. There was no noticeable decrease in rupture temperature with increasing fluence in lower hoop stress region($$sim$$100MPa). (2)The rupture temperature of the irradiated 15Cr-20Ni fuel cladding of MFA-2 was almost as same as that of as-received cladding for the hoop stress range up to about 200MPa. The rupture temperature did not decrease significantly with fluence. (3)The rupture temperature of the irradiated PNC316 cladding tested at hoop stress 69MPa, which was the design hoop stress for MONJU fuel, was 1055.6$$^{circ}$$C. This suggested that the design cladding maximum temperature limit for MONJU (830$$^{circ}$$C) was conservative. (4)There was no obvious relation between rupture temperature, swelling and microstructural change during transient heating under the irradiation conditions examined.

JAEA Reports

Analysis of Cm contained in irradiated fuel of experimental fast reactor "JOYO"; Development of the analytical technique and measurement of Cm

Osaka, Masahiko; Koyama, Shinichi; Mitsugashira, Toshiaki; Morozumi, Katsufumi; Namekawa, Takashi

JNC-TN9400 2000-058, 49 Pages, 2000/04

JNC-TN9400-2000-058.pdf:1.22MB

The analytical technique for Cm contained in a MOX FUEL was developed and analysis of Cm contained in irradiated fuel of experimental fast reactor "JOYO" was carried out, to contribute to evaluation of transmutation characteristics of MA nuclide in the fast reactor. The procedure of ion-exchange separation of Cm with nitric acid-methanol mixed media essential for the isotopic analysis in irradiated MOX fuel was adopted considering for being rapid and easy. The fundamental test to grasp separation characteristics of this procedure, such as Cm elution position and separation capacity between Cm and Am or Eu, was carried out. ln applying this procedure to the analysis of Cm contained in actual specimen, separation condition was evaluated and optimized, and the procedure consist of impurity removal and Am removal process was devised. This procedure resulted in high recovery rate of Cm and high removal rate of Am and impurity which becomes a problem in sample handling and mass-spectrometry such as Eu and Cs. The Cm separation test from irradiated MOX fuel was carried out using this technique, and Cm isotopic ratio analysis was enabled. The analytical technique for Cm contained in irradiated MOX fuel was established using the procedure of ion-exchange separation with nitric acid-methanol mixed media. The analysis of Cm contained in irradiated MOX fuel of experimental fast reactor "Joyo" was carried out. As a result, it was revealed from measured data that Cm content rate was 1.4$$sim$$ 4.0$$times$$lO$$^{-3}$$ atom%, small amount of $$^{247}$$Cm was generated and Cm isotopic ratio was constant above burn-up 60GWd/t.

JAEA Reports

JOYO MK-II core plant characteristics test data

JNC-TN9410 2000-010, 72 Pages, 2000/03

JNC-TN9410-2000-010.pdf:2.14MB

The experimental fast reactor JOYO served as the MK-II irradiation bed core for testing fuel and material for FBR development for 16 years from 1982 to 1997. During the MK-II core operation, extensive data were accumulated from the plant characteristic tests. Tests conducted at JOYO included operating characteristic tests for confirming operational safety, performance tests for confirming design performance of the MK-II core, and special tests for research and development ofthe plant. In this report, the outline and the results of each test item are shown. These test data can be provided by the magnet-optical disk.

JAEA Reports

JOYO coolant sodium and cover gas purity control database (MK-II core)

; ; Saikawa, Takuya*; Sukegawa, Kazuya*

JNC-TN9410 2000-008, 66 Pages, 2000/03

JNC-TN9410-2000-008.pdf:1.39MB

The experimental fast reactor "JOYO" served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, impurities concentrations in the sodium and the argon gas were determined by 67 samples of primary sodium, 81 samples of secondary sodium, 75 samples of primary argon gas, 89 samples of secondary argon gas (the overflow tank) and 89 samples of secondary argon gas (the dump tank). The sodium and the argon gas purity control data were accumulated from in thirty-one duty operations, thirteen special test operations and eight annual inspections. These purity control results and related plant data were compiled into database, which were recorded on CD-ROM for user convenience. Purity control data include concentration of oxygen, carbon, hydrogen, nitrogen, chlorine, iron, nickel and chromium in sodium, concentration of oxygen, hydrogen, nitrogen, carbon monoxide, carbon dioxide, methane and helium in argon gas with the reactor condition.

JAEA Reports

Development of JOYO plant operation management expert Tool (JOYPET)

; Terano, Toshihiro; ; ;

JNC-TN9410 2000-004, 30 Pages, 2000/03

JNC-TN9410-2000-004.pdf:0.86MB

The Operation and Maintenance Support Systems for JOYO are being developed, with the aim of ensuring the stable and safe operation of JOYO and improving operational reliability of future FBR plants. Plant Operation Management Expert Tool named JOYPET had been developed as one of the Operation and Maintenance Support Systems, which helps plant operation management. The following functions were developed and applied. (1)Papers management (Plant status management) function for maintenance activities (2)Isolation management support function for plant operation (3)Automatically drawing function of plant operation schedule (4)Isolation judgment function for plant operation By use this system, the plant management of JOYO was able to improved reliability and reduced manpower.

JAEA Reports

Development of a standard database for FBR core nuclear design (XI); Analysis of the experimental fast reactor "JOYO" MK-I start up test and oparation data

; Numata, Kazuyuki*

JNC-TN9400 2000-036, 138 Pages, 2000/03

JNC-TN9400-2000-036.pdf:10.16MB

Japan Nuclear Cycle Development lnstitute (JNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were renected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. ln this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor "JOYO" MK-l core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. 0n the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of "JOYO" MK-l core in comparison with ZPPR-9 core of JUPITER experiments.

Journal Articles

Thermal hydraulic characteristics during ingress of coolant and loss of vacuum events in fusion reactors

Takase, Kazuyuki; Kunugi, Tomoaki*; Seki, Yasushi; Akimoto, Hajime

Nuclear Fusion, 40(3Y), p.527 - 535, 2000/03

 Times Cited Count:8 Percentile:71.52(Physics, Fluids & Plasmas)

no abstracts in English

170 (Records 1-20 displayed on this page)