DEMO plant design beyond ITER

Konishi, Satoshi; Nishio, Satoshi; Tobita, Kenji; DEMO Design Team

Fusion Engineering and Design, 63-64, p.11 - 17, 2002/12

https://doi.org/10.1016/S0920-3796(02)00264-8

The first fusion power plant DEMO must have some reality that ITER and other facilities in the same period are expected to prove its feasibility. The DEMO should also be so attractive and advanced that the future society would be interested in constructing based on its concept. The present DEMO plant concept intends to satisfy these antagonistic requirements assuming construction in 2030s immediately after successful completion of fundamental ITER mission. A steady tokamak is minimized to have 5.8m of major radius with 2.3GW with Q exceeds 30. Modestly ambitious plasma parameters are chosen. Technology improvement is assumed to make maximum 20 T magnet, metal first wall and super critical water cooled ITER-like blanket modules feasible. Tritium inventory is reduced to 1kg with improved safety system concept. This conceptual design identifies various technical issues that are expected to be solved by intensive R&D efforts during ITER period, and indicates a possible step immediately after ITER.

Current status and future plan for thermaI striping investigations at JNC

; kasahara, Naoto; ; ; Kamide, Hideki

JNC TN9400 2000-010, 168 Pages, 2000/02

JNC-TN9400-2000-010.pdf:8.78MB

Thermal striping is significant issue of the structural integrity, where the hot and cold fluids give high cycle fatigue to the structure through the thermal stress resulted from the time change of temperatur distibution in the structure. In the sodium cooled fast reactor, temperature change in fluid easily transfers to the structure because of the high thermal conductivity of the sodium. It means that we have to take care of thermal striping, The thermal striping is complex phenomena between the fluid and structure engineering fields. The investigations of thermal striping are not enough to evaluate the integrity directly. That is the fluctuation intensity at the structure surface is assumed to be temperature difference between source fluids (upstream to the mixing region) as the maximum value in the design. 0therwise, the design conditions are defined by using a mockup experiment and large margin of temperature fluctuation intensity. Furthermore, such evaluation manners have not yet been considered as a design rule. Transfer mechanism of temperature fluctuation from fluid to structure has been investigated by the authors on the view points of the fluid and structure. Attenuation of temperature fluctuation was recognized as a dominant factor of thermal fatigue. We have devdoped a numerical analysis system which can evaluate thermal fatigue and crack growth with consideration of the attenuation of temperature fluctuation in fluid, heat transfer, and structure. This system was applied to a real reactor and the applicability was confirmed. Further verification is planned to generalize the system. For the higher cost performance of the fast reactor, an evaluation rule is needed, which can estimate thermal loading with attenuation and can be applied to the design. An idea of the rule is proposed here. Two methods should be prepared; one is a precise evaluation method where mechanism of attenuation is modeled, and the other is simple evaluation method where ...

Annual report of Naka Fusion Research Establishment from April 1, 1998 to March 31, 1999

Naka Fusion Research Establishment

JAERI-Review 99-022, p.118 - 0, 1999/09

JAERI-Review-99-022.pdf:9.51MB

no abstracts in English

Recent activities on subchannel analysis at JAERI

Okubo, Tsutomu; Araya, Fumimasa; Iwamura, Takamichi; Kusunoki, Tsuyoshi

Fourth Int. Seminar on Subchannel Analysis (ISSCA-4), p.267 - 286, 1997/00

no abstracts in English

None

; Kobayashi, Hiroaki; Masaki, Toshio; *; Kawamura, Kazuhiro; ;

PNC TN8440 94-028, 173 Pages, 1994/06

None

Proceedings of the 8th Topical Meeting on Nuclear Code Development; November 11-12,1992,JAERI Tokai,Japan

Research Committee on Reactor Physics;

JAERI-M 93-033, 246 Pages, 1993/03

JAERI-M-93-033.pdf:7.78MB

no abstracts in English

Information retrieval system of Nuclear Power Plant Database(PPD); Users guide

Izumi, Fumio; ; Kobayashi, Kensuke

JAERI-M 90-218, 67 Pages, 1990/12

JAERI-M-90-218.pdf:1.92MB

no abstracts in English

Effects of specimen volume on tensile strength for nuclear-grade

; ;

Tanso, 1985(120), p.45 - 47, 1985/00

https://doi.org/10.7209/tanso.1985.45

no abstracts in English

Annual Report of Division of Thermonuclear Fusion Research and Division of Large Tokamak Development for the Period of April 1, 1976 to March 31, 1977

;

JAERI-M 7479, 334 Pages, 1978/02

JAERI-M-7479.pdf:9.82MB

no abstracts in English

Fusion structural material development in view of DEMO design requirement

Tanigawa, Hiroyasu

no journal, , 

The fusion structural materials, especially those in blanket system, will be subjected to high heat loads and high flux 14 MeV fusion neutrons, and will experience thermal stresses, substantial magnetic forces in case of plasma disruptions, and high pressure loads in the case of a coolant leak. For this use, the structural material must have an adequate database including irradiation database to assure the mechanical properties, and to define the limits for its usage, along with sound technical bases on full scale manufacturing and joining technologies. In this talk, the strategy to develop fusion structural material in view of design requirement is discussed in case of structural material development for fusion blanket system.