Development of ITER divertor vertical target with annular flow concept, 2; Development of brazing technique for CFC/CuCrZr joint and heating test of large-scale mock-up

Ezato, Koichiro; Dairaku, Masayuki; Taniguchi, Masaki; Sato, Kazuyoshi; Suzuki, Satoshi; Akiba, Masato; Ibbott, C.*; Tivey, R.*

Fusion Science and Technology, 46(4), p.530 - 540, 2004/12

https://doi.org/10.13182/FST04-A588

The first fabrication and heating test of a large-scale CFC monoblock divertor mock-up using annular flow concept have been performed to demonstrate its manufacturability and thermo-mechanical performance. Prior to the fabrication of the mock-up, brazed joint tests between the CFC monoblock and the CuCrZr tube have been carried out to find the suitable heat treatment mitigating loss of the high mechanical strength of the CuCrZr material. Basic mechanical examination on CuCrZr undergoing the brazing heat treatment and FEM analyses are also performed to support the design of the mock-up. High heat flux tests on the large-scale divertor mock-up have been performed in an ion beam facility. The mock-up has successfully withstood more than 1,000 thermal cycles of for 15 s and 3,000 cycles more than for 15 s, which simulates the heat load condition of the ITER divertor. No degradation of the thermal performance of the mock-up has been observed throughout the thermal cycle test.

Structural integrity assessment of helium component during safety demonstration test using HTTR, 1 (Contract Research)

Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki; Tachibana, Yukio

JAERI-Tech 2004-045, 67 Pages, 2004/04

JAERI-Tech-2004-045.pdf:4.74MB

Safety demonstration tests using the HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactors. In the safety demonstration tests, the coolant flow reduction test by tripping one or two out of three gas circulators is being performed between FY2002 and FY 2005 and by tripping all the three gas circulators will be conducted after FY2006. This paper describes the structural integrity assessment of the primary pressurised water cooler after one and two gas circulators run down. Also, the possibility of natural convection in the primary coolant after all the three gas circulator stopped was evaluated by the operation data of the reactor-scram test performed during the rise-to-power tests.

Safety demonstration tests using High Temperature Engineering Test Reactor (HTTR)

Tachibana, Yukio; Nakagawa, Shigeaki; Iyoku, Tatsuo

Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 8 Pages, 2003/09

no abstracts in English

Cause and countermeasure for heat up of HTTR core support plate at power rise tests

Fujimoto, Nozomu; Takada, Eiji*; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatsuo

JAERI-Tech 2001-090, 69 Pages, 2002/01

JAERI-Tech-2001-090.pdf:7.88MB

HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate showed higher results than expected value at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in a core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses.

Core void fraction distribution under high-temperature high-pressure boil-off conditions; Experimental study with two-phase flow test facility(TPTF)

; Kumamaru, Hiroshige; Murata, Hideo; Anoda, Yoshinari; Kukita, Yutaka

JAERI-M 93-200, 56 Pages, 1993/10

JAERI-M-93-200.pdf:1.85MB

no abstracts in English

BETHSY/LSTF counterpart test on natural circulation in a pressurized water reactor

P.Bazin*; R.Deruaz*; Yonomoto, Taisuke; Kukita, Yutaka

ANS Proc. of the 1992 National Heat Transfer Conf., p.301 - 308, 1992/00

no abstracts in English

Construction of in-core structure test section in HENDEL, (II); Testing devices and their performance test

; ; ; ; ; ; *;

Nihon Genshiryoku Gakkai-Shi, 30(5), p.427 - 433, 1988/05

https://doi.org/10.3327/jaesj.30.427

no abstracts in English

BWR Recirculation Loop Discharge Line Break LOCA Tests with Break Areas 50% and 100% Assuming HPCS Failure at ROSA-III Test Facility

; Tasaka, Kanji; ; ; ; ; ; *;

JAERI-M 85-037, 224 Pages, 1985/03

JAERI-M-85-037.pdf:5.99MB

no abstracts in English

Core Hydrodynamic Tests and Determination of Core Flow of JRR-3 to Be Renewed ata 20MWt

Sudo, Yukio; ; *;

JAERI-M 84-119, 108 Pages, 1984/06

JAERI-M-84-119.pdf:2.16MB

no abstracts in English

Evaluation Report on CCTF Core-I Reflood Tests Cl-16(Run 25),Cl-21(Run 40)and Cl-22(Run 41); Comparison of Rresults Between FLECHT Coupling Tests and FLECHT-SET Test

Murao, Yoshio; ;

JAERI-M 83-065, 113 Pages, 1983/05

JAERI-M-83-065.pdf:2.07MB

no abstracts in English