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Kato, Chiaki; Yano, Masaya*; Kiuchi, Kiyoshi; Sugimoto, Katsuhisa*
Corrosion Engineering, 52(1), p.53 - 67, 2003/01
The effects of heat-transfer on the corrosion of zirconium was examined in boiling nitric acid solutions with various concentrations. Corrosion mass losses and electrochemical polarization curves were measured on the heat-transfer and isothermal surfaces in the solutions. It was found that the corrosion rate of zirconium was higher on the heat-transfer surface than that on the isothermal surface. The rate increased with increasing nitric acid concentration and solution temperature. The increased oxidization potential on the heat-transfer surface is attributed to the reduction of nitrous acid concentration by the thermal decomposition on the surface and the removal of the decomposition product from solution by boiling bubbles. The redox potential of 12 mol/dm
nitric acid on a boiling heat-transfer surface was very close to the breakdown potential of primary passivity of zirconium. This suggests the initiation of SCC on a boiling heat-transfer surface in a nuclear fuel reprocessing.
Kato, Chiaki; Yano, Masaya*; Kiuchi, Kiyoshi; Sugimoto, Katsuhisa*
Zairyo To Kankyo, 52(1), p.35 - 43, 2003/01
The effects of heat-transfer on the corrosion of zirconium was examined in boiling nitric acid solutions with various concentrations. Corrosion mass losses and electrochemical polarization curves were measured on the heat-transfer and isothermal surfaces in the solutions. It was found that the corrosion rate of zirconium was higher on the heat-transfer surface than that on the isothermal surface. The rate increased with increasing nitric acid concentration and solution temperature. The increased oxidization potential on the heat-transfer surface is attributed to the reduction of nitrous acid concentration by the thermal decomposition on the surface and the removal of the decomposition product from solution by boiling bubbles. The redox potential of 12 mol/dm3 nitric acid on a boiling heat-transfer surface was very close to the breakdown potential of primary passivity of zirconium. This suggests the initiation of SCC on a boiling heat-transfer surface in a nuclear fuel reprocessing.
Doi, Masamitsu; Kiuchi, Kiyoshi; Yano, Masaya*; Sekiyama, Yoshio*
JAERI-Research 2001-020, 17 Pages, 2001/03
JAERI-Research-2001-020.pdf:0.7MB
no abstracts in English
Saburi, Tei; Ogawa, Hiroaki; Ueda, Satoshi*; Kiuchi, Kiyoshi
JAERI-Tech 2000-057, 23 Pages, 2000/10
JAERI-Tech-2000-057.pdf:5.28MB
no abstracts in English
Sakai, Takaaki; *; Ohshima, Hiroyuki; Yamaguchi, Akira
JNC TN9400 2000-033, 94 Pages, 2000/04
JNC-TN9400-2000-033.pdf:4.36MB
The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. ln this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube fairer accidents in a steam generator. ln this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in "Equivalent plant" with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. ln conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to conform the heat transfer reduction by the oxidize film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance.
; Kamide, Hideki;
PNC TN9410 98-083, 118 Pages, 1998/07
Large-scaled thermohydraulic tests are planned for new key technologies in the heat transport systems of a demonstration fast reactor. The test facility is consisted of components from a reactor vessel to a steam generator (SG). Basic design of the large-scaled thermohydraulic test facility is 1/3 scale of the demonstration fast reactor with two primary cooling loops and two into one secondary loop. The secondary piping length of the test facility is longer than the 1/3 scale of the demonstration fast reactor. The tests facility has the branch and junction of the secondary piping because of two primary loops and one SG. There is a possibility of flow and temperature unbalance if a buoyancy force were large and pressure loss were small. Therefore, dynamics analyses of the thermal transition tests had been done in which the secondary piping length. To examine the unbalance occurred or not, the natural circulation analysis had been performed providing different heat transfer area of the IHX or presser loss of the primary loop between A loop and B loop. It was shown from the analyses that the temperature response during the transition was delayed in the test model compared to the real reactor. Main cause of the delay was due to the real scaled SG. Other parameters, the length of piping etc., were not very influential to the response. The analysis such predicted that there wasn't large difference of global behaviors between the loops. Therefore, it was shown that there would be no problem, if the difference were made between the loops due to a manufacturing error.
Takase, Kazuyuki; Kunugi, Tomoaki*; Akimoto, Hajime
Proc. of 6th Int. Conf. on Nucl. Eng. (CD-ROM), 12 Pages, 1998/00
no abstracts in English
Hishida, Makoto
J. Enhanced Heat Transfer, 3(3), p.187 - 200, 1996/00
no abstracts in English
*; Kiuchi, Kiyoshi; *;
Proc. of the Int. Symp. on Material Chemistry in Nuclear Environment, p.469 - 477, 1992/00
no abstracts in English
Hishida, Makoto; Takase, Kazuyuki
Proceedings of the ASME-JSME Thermal Engineering Joint Conference : Reno, Nevada, March 17-22, 1991, p.103 - 110, 1991/00
no abstracts in English
Hishida, Makoto
Nihon Kikai Gakkai Rombunshu, B, 56(524), p.1107 - 1112, 1990/04
no abstracts in English
Hishida, Makoto
Nihon Kikai Gakkai Rombunshu, B, 55(518), p.3166 - 3171, 1989/10
no abstracts in English
Okamoto, Yoshizo
Nihon Kikai Gakkai Rombunshu, 21(224), p.624 - 631, 1965/00
no abstracts in English
