Current status of experimental study and device modifications in JT-60U

Kurihara, Kenichi; JT-60 Team

Proceedings of 21st IEEE/NPSS Symposium on Fusion Engineering (SOFE 2005) (CD-ROM), 6 Pages, 2005/09

Since JT-60 is expected to explore more advanced operation scenarios, the discharge pulse length and the duration time of additional NBI/RF heating were extended to 65 s and 30 s/60 s, respectively, in 2003. The experimental campaign in 2003-2004 has ended up with the following significant results: (a) The high bootstrap current ratio of 75 % was sustained for 7.4 s in an R/S plasma. (b) The quasi-steady state beta value was increased to 3.0 for a pulse of 6.2 s with NTM suppression by ECCD, etc. For further exploration toward high performance plasmas, the following modifications have been conducted: (1) To minimize the power loss at the region of toroidal field ripple, the 8Cr ferritic steel tiles are being equipped on the first wall of the vacuum vessel. (2) A new current profile reproduction method will be installed in the plasma control system. In the symposium, the current status of plasma experimental study will be presented together with on-going device modifications in JT-60.

None

*; *; *

JNC TJ8420 2000-003, 99 Pages, 2000/03

JNC-TJ8420-2000-003.pdf:5.47MB

no abstracts in English

Comparison of edge neutral effect on the condition of H-mode transition before and after the modified divertor in JT-60U

; Fukuda, Takeshi; Kamada, Yutaka; Takenaga, Hidenobu; Takizuka, Tomonori; Mori, Masahiro; Fujita, Takaaki; JT-60 Team

Plasma Physics and Controlled Fusion, 40(5), p.713 - 716, 1998/00

https://doi.org/10.1088/0741-3335/40/5/029

no abstracts in English

None

; ;

PNC TN9420 97-001, 82 Pages, 1996/10

PNC-TN9420-97-001.pdf:2.58MB

None

Operation and maintenance experience on the fuel handling systems and storage facilities of "MONJU", 1

; ; Yamada, Takeshi; ; ; ; Kaito, Yasuaki; ; Kotaka, Yoshinori; ; et al.

PNC TN2410 96-005, 339 Pages, 1996/03

PNC-TN2410-96-005.pdf:14.53MB

Construction of the 'Monju' fuel handling systems was completed in April, 1992. From March 1991 to August 1992, pre-commissioning tests were carried out. In December 1992, all the systems of Monju were transfered to PNC, and commissioning tests and reactor physics tests, were started. For the first time, during these physics tests, the fuel handling systems were operated for one of the commissioning tests 'Loading to Criticality', without significant problems. 168 fuel sub-assemblies were loaded into the core and the first criticality was achieved on 5th April 1994. The fuel handling systems continued in operation for the 'Loading to Full Size of the Core', power distribution test and for cleaning discharged dummy sub-assemblies. To keep these fuel handling systems working somothly and satisfactorily annual maintenance has been carried out since 1992. This paper describes the operation and maintenance experience of fuel handling systems after the pre-commissioning tests and future study items for system reliability improvement.

Recent topics on probabilistic safety assessment

Sobajima, Makoto; Muramatsu, Ken

Genshiryoku Kogyo, 42(9,10), p.7 - 13, 1996/00

http://ci.nii.ac.jp/naid/40001074529

no abstracts in English

None

Yoshida, Mamoru; ; Shobu, Nobuhiro; Aihara, Nagafumi;

PNC TN9700 95-001, 804 Pages, 1995/08

PNC-TN9700-95-001.pdf:30.97MB

no abstracts in English

None

; Miyazaki, Hitoshi; ; Tanimoto, Kenichi; Terunuma, Seiichi

PNC TN9420 94-010, 103 Pages, 1994/04

PNC-TN9420-94-010.pdf:2.89MB

None

Preliminary design for reconstruction of SWAT-3 facility

*; *; *; *; *; *; *

PNC TJ9164 94-006, 133 Pages, 1994/03

PNC-TJ9164-94-006.pdf:3.4MB

This report gives an applicability of SWAT-3 facility and contents of the reconstruction in order to confirm a DBL (Design Basis Leak) for the demonstration reactor SG. (1).Test Cndition and test case. Evaluation of the wall temperature for adjacent heat transfer tubes under the sodium-water reaction event was performed. (a)As the effect of tube rupture due to overheating, failure of upper part of the helical coil was severer than one of the lower part. (b)The wall temperature depends on the water side condition. (c)Reference test condition, whici is water leak rale about 1 kg/s, failure of upper part of the helical coil and 30% partial load, was selected. A total of ten test cases were decided. (2).System and Components Design. (a)Large leak sodium-water reaction analyses including water injection rate analysis and quasi-steady pressure analysis were performed. The maximum water leak rate of 1 DEG was 7.2 kg/s and the water leak rate at the quasi-steady state was 3.1 kg/s. The maximum pressure was 18.1kg/cma at the piping between the reaction vessel and IHX, the pressure was within the design condition of SWAT-3 facility. (b)Based on the results of the large leak sodium-water reaction analyses, a reaction vessel, water heaters and a dump tank were designed and their design specification were clarified. The reaction vessel was a scale of one third of the demonstration reactor SG and it was designed to be able to conduct the water injection test twice with one test unit. (c)A system and piping diagram, and many kinds of list (Piping list, Valve list, instrumentlist) were made up. (3).Reconstruction scope and arrangement plan. The reconstruction scope and a layout for the components and piping were clarfied and the arrange ment plans were made up. (4)Reconstruction period. The recoastruction period and man power for the design, fabrication, inspection and installation were studied and the reconstruction schedule was made up.

None

PNC TN1700 93-005, 139 Pages, 1993/01

PNC-TN1700-93-005.pdf:2.94MB

no abstracts in English