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Journal Articles

Applicability study of nuclear graphite material IG-430 to VHTR

Osaki, Hirotaka; Shimazaki, Yosuke; Sumita, Junya; Shibata, Taiju; Konishi, Takashi; Ishihara, Masahiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05

For the design on the VHTR graphite components, it is desirable to employ graphite material with higher strength. IG-430 graphite has been developed as an advanced candidate for VHTR. However, the new developed IG-430 does not have enough databases for the design of HTGR. In this paper, the compressive strength (Cs) of IG-430, one of important strengths for design data, is statistically evaluated. The component reliability is evaluated based on the safety factors defined by the graphite design code, and the applicability as the VHTR graphite material is discussed. It was found that IG-430 has higher strength (about 11%) and lower standard deviation (about 27%) than IG-110 which is one of traditional graphites used for HTGR, because the crack in IG-430 would not easy to propagate rather than IG-110. Since fracture probability for IG-430 is low, the higher reliability of core-component will be achieved using IG-430. It is expected that IG-430 is applicable for VHTR graphite material.

Journal Articles

An Approach for development of technical structural standard in ITER

Nakahira, Masataka; Takeda, Nobukazu

Hozengaku, 4(4), p.47 - 52, 2006/01

https://ci.nii.ac.jp/naid/40007103745

The technical structural standard for ITER (International Thermonuclear Experimental Fusion Reactor) should be innovative because of their quite different features of safety and mechanical components from nuclear fission reactors, and the necessity of introducing several new fabrication and examination technologies. Recognizing the international importance of Fusion Standard, Japan and ASME has started the cooperation development of the Fusion Standard. This paper shows the special features of ITER from view points of safety, design and fabrication, and proposes approach for development of the fusion standard.

JAEA Reports

Structural analysis of support structure for ITER vacuum vessel

Takeda, Nobukazu; Omori, Junji*; Nakahira, Masataka

JAERI-Tech 2004-068, 27 Pages, 2004/12

JAERI-Tech-2004-068.pdf:7.68MB

ITER vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed. This independent concept has two advantages: (1) thermal load due to the temperature deference between VV and the lower temperature components such as TF coil becomes lower and (2) the other components such as TF coil is categorized as a non-safety component because of its independence from VV. Stress analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support.

Journal Articles

Design and structural analysis of support structure for ITER vacuum vessel

Takeda, Nobukazu; Omori, Junji*; Nakahira, Masataka; Shibanuma, Kiyoshi

Journal of Nuclear Science and Technology, 41(12), p.1280 - 1286, 2004/12

https://doi.org/10.3327/jnst.41.1280

Times Cited Count：3 Percentile：23.52(Nuclear Science & Technology)

ITER vacuum vessel (VV) is a safety component confining radioactive materials. An independent VV support structure located at the bottom of VV lower port is proposed as an alternative concept, which is deferent from the current reference, i.e., the VV support is directly connected to the toroidal coil (TF coil). This independent concept has two advantages comparing to the reference one: (1) thermal load becomes lower and (2) the TF coil is categorized as a non-safety component. Stress Analyses have been performed to assess the integrity of the VV support structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as an alternative VV support.

JAEA Reports

Revised version of tokamak transient simulation code SAFALY, 2

Senda, Ikuo*; Fujieda, Hirobumi; Neyatani, Yuzuru; Tada, Eisuke; Shoji, Teruaki

JAERI-Data/Code 2003-012, 73 Pages, 2003/07

JAERI-Data-Code-2003-012.pdf:3.45MB

The tokamak transient simulation code, named SAFALY, was revised recently and the sensitivity analyses on the parameters in the code were carried out. This report is composed of two volumes. The formulation and the parameters in modeling the plasma and in-vessel components are described in the first volume. In this second volume, the results of the sensitivity studies are reported. The sensitivity studies were performed in two steps. In the first step, the responses of plasmas in the occurrence of plasma disturbances were analyzed for various initial conditions. For each disturbance, the initial condition of the plasma, which gave the largest increase of the fusion power, was identified. In the second step, by using initial conditions derived in the first step, the sensitivities of plasma reactions with respect to variation of the parameters in SAFALY code were analyzed. In the analyses, the increase of the fueling, the increase of the plasma confinement improvement factor and the increase of the auxiliary heating power were considered as plasma disturbances.

JAEA Reports

Revised version of tokamak transient simulation code SAFALY, 1

Senda, Ikuo*; Fujieda, Hirobumi; Neyatani, Yuzuru; Tada, Eisuke; Shoji, Teruaki

JAERI-Data/Code 2003-008, 37 Pages, 2003/06

JAERI-Data-Code-2003-008.pdf:1.58MB

Tokamak transient simulation code, named SAFALY, was revised. SAFALY code has been developed to simulate transient events in Tokamaks. Modeling of the plasma and algorithms of the simulation were revised. The code was also modified to deal with the variation of the plasma current. The code was improved to allow flexible modeling of in-vessel components. The data transfer between SAFALY and related codes was arranged to prepare data required in analyses with SAFALY, such as the distributions of heat/neutron loads and the radiation form factor between in-vessel components. The report is composed of two volumes. The formulation and the parameters in modeling plasma and in-vessel components are described in this first volume. Examples of simulation results, using the design of ITER-FDR in 2001, are presented and general properties of plasmas' responses with respect to perturbations are discussed. The results of the sensitivity studies with respect to simulation parameters and initial conditions will be reported in the second volume.

JAEA Reports