Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Aihara, Jun; Ueta, Shohei; Goto, Minoru; Inaba, Yoshitomo; Shibata, Taiju; Ohashi, Hirofumi
JAEA-Technology 2018-002, 70 Pages, 2018/06
HTFP code is code for calculation of additional release amount of fission product (FP) from fuel rod in high temperature gas-cooled reactor (HTGR) after stop of fission. Minory changed Fornax-A code also can calculate that. Therefore, release behavior of Cs calculated with HTFP code was compared with that calculated with minory modified FORNAX-A code in this report. Release constants of Cs evaluated with minory modified FORNAX-A code are rather different from default values for HTFP code.
Nomoto, Yasunobu; Aihara, Jun; Nakagawa, Shigeaki; Isaka, Kazuyoshi; Ohashi, Hirofumi
JAEA-Data/Code 2015-008, 39 Pages, 2015/06
HTFP is a calculation code for amount of additionally released fission product (FP) from fuel rods of pin-in-type according to transient of core temperature at the accident of high temperature gas-cooled reactors (HTGRs). This code analyzes FP release inventory from core according to the transient of core temperature at the accident as an input data and considering FP release rate from a fuel compact and a graphite sleeve and radioactive decay of FP. This report describes the outline of HTFP code and its input data. The computed solutions using the HTFP code were compared to those of HTCORE code, which was used for the design of the High Temperature Engineering Test Reactor (HTTR) to validate the analysis models of the HTFP code. The comparison of HTFP code results with HTCORE code results showed the good agreement.
Tashiro, Shinsuke; Uchiyama, Gunzo; Amano, Yuki; Abe, Hitoshi; Yamane, Yuichi; Yoshida, Kazuo
Nuclear Technology, 190(2), p.207 - 213, 2015/05
Times Cited Count:8 Percentile:53.32(Nuclear Science & Technology)The release behavior of radioactive materials from high active liquid waste (HALW) has been investigated under boiling accident conditions. Results of the experiment using a nonradioactive simulated HALW found Ru to be a volatile element under the accident conditions and to be released into the gas phase in the form of both mist and gas. The Ru release rate and the apparent Ru volatilization rate constant were obtained under the boiling conditions of simulated HALW. The other fission product elements such as Cs were found to be nonvolatile and to be released into the gasphase in the form of mist. The mist size distribution near the surface of the simulated HALW in the reactor vessel was found to range from 0.05 to 20 m with a peak diameter of
2
m.
Tanaka, Kosuke*; Maeda, Koji*; Katsuyama, Kozo*; Inoue, Masaki*; Iwai, Takashi; Arai, Yasuo
Journal of Nuclear Materials, 327(2-3), p.77 - 87, 2004/05
no abstracts in English
Inoue, Masaki*; Iwai, Takashi; Arai, Yasuo; Asaga, Takeo*
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.1694 - 1703, 2003/11
no abstracts in English
Arai, Yasuo; Iwai, Takashi; Nakajima, Kunihisa; Nagashima, Hisao; Nihei, Yasuo; Katsuyama, Kozo*; Inoue, Masaki*
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.1686 - 1693, 2003/00
no abstracts in English
Fuel Safety Research Laboratory
JAERI-Conf 2002-009, 491 Pages, 2002/08
Fuel Safety Research Specialists' Meeting, which was organized by Japan Atomic Energy Research Institute, was held on March 4-5, 2002 at JAERI in Tokai Establishment. Purposes of the Meeting are to exchange information and views on LWR fuel safety topics among the specialist participants from domestic and foreign organizations, and to discuss the recent and future fuel research activities in JAERI. In the Meeting, presentations were given and discussions were made on general report of fuel safety research activities, fuel behaviors in normal operation and accident conditions, FP release behaviors in severe accident conditions, and JAERI's "Advanced LWR Fuel Performance & Safety Research Program". A poster exhibition was also carried out. The Meeting significantly contributed to planning future program and cooperation in fuel research. This proceeding integrates all the pictures and papers presented in the Meeting.
Fuel Safety Research Laboratory
JAERI-Conf 2001-010, 303 Pages, 2001/09
The 24th NSRR Technical Review Meeting was held at Tranomon Pastoral, Tokyo, on November 13 and 14, 2000. The purpose of the meeting was to present and discuss the recent progress of the NSRR program and other LWR fuel safety researches at JAERI. Twenty-one papers, including five by foreign institutes, were presented and discussed regarding fuel behavior during normal operation, reactivity initiated accident (RIA) and loss-of-coolant accident (LOCA) and FP release behavior during severe accident. The meeting was a great help in planning future research and promoting research cooperation. This proceeding contains the papers presented in the meeting.
Suzuki, Motoe
JAERI-Data/Code 2000-030, 280 Pages, 2000/09
no abstracts in English
Suzuki, Motoe; Saito, Hioraki*
HPR-349, 12 Pages, 1998/00
no abstracts in English
Suzuki, Motoe; Saito, Hioraki*
JAERI-Data/Code 97-046, 210 Pages, 1997/11
no abstracts in English
Suzuki, Motoe; Saito, Hioraki*
JAERI-Data/Code 97-043, 235 Pages, 1997/11
no abstracts in English
Suzuki, Motoe; Saito, Hioraki*
JAERI-Data/Code 97-010, 221 Pages, 1997/03
no abstracts in English
Suzuki, Motoe; Saito, Hioraki*
JAERI-Data/Code 94-011, 178 Pages, 1994/09
no abstracts in English
Matsumoto, Mikio; Endo, Yasuichi; ; Itabashi, Yukio; ; Yokouchi, Iichiro; Ando, Hiroei
JAERI-M 92-212, 62 Pages, 1993/01
no abstracts in English
Yanagisawa, Kazuaki
Journal of Nuclear Science and Technology, 29(9), p.909 - 918, 1992/09
no abstracts in English
; Saito, Minoru; Oyamada, Rokuro; ; ; Saito, Junichi; Iwai, Takashi; ; Nakagawa, Tetsuya
Nucl. Saf., 33(3), p.334 - 343, 1992/07
no abstracts in English
; ; Saito, Junichi; Komukai, Bunsaku; Ando, Hiroei; ; ; ; Saito, Minoru;
Proc. on 12th Int. Meeting,Reduced Enrichment for Research and Test Reactors, p.269 - 280, 1991/00
no abstracts in English
Yanagisawa, Kazuaki
Nihon Genshiryoku Gakkai-Shi, 32(4), p.385 - 394, 1990/04
Times Cited Count:1 Percentile:19.47(Nuclear Science & Technology)no abstracts in English
; ; E.Kolstad*
Nihon Genshiryoku Gakkai-Shi, 28(7), p.641 - 657, 1986/00
Times Cited Count:4 Percentile:47.80(Nuclear Science & Technology)no abstracts in English