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JAEA Reports

Comparison between HTFP code and minory changed FORNAX-A code

Aihara, Jun; Ueta, Shohei; Goto, Minoru; Inaba, Yoshitomo; Shibata, Taiju; Ohashi, Hirofumi

JAEA-Technology 2018-002, 70 Pages, 2018/06

JAEA-Technology-2018-002.pdf:1.46MB

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.

JAEA Reports

HTFP for calculation of amount of additionally released fission products from fuel rods of pin-in-block-type high temperature gas-cooled reactors during accident

Nomoto, Yasunobu; Aihara, Jun; Nakagawa, Shigeaki; Isaka, Kazuyoshi; Ohashi, Hirofumi

JAEA-Data/Code 2015-008, 39 Pages, 2015/06

JAEA-Data-Code-2015-008.pdf:10.32MB

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.

Journal Articles

Fission gas release and swelling in uranium-plutonium mixed nitride fuels

Tanaka, Kosuke*; Maeda, Koji*; Katsuyama, Kozo*; Inoue, Masaki*; Iwai, Takashi; Arai, Yasuo

Journal of Nuclear Materials, 327(2-3), p.77 - 87, 2004/05

https://doi.org/10.1016/j.jnucmat.2004.01.002

no abstracts in English

Journal Articles

Analysis of mechanical load on cladding induced by fuel swelling during power ramp in high burn-up rod by fuel performance code FEMAXI-6

Suzuki, Motoe; Uetsuka, Hiroshi; Saito, Hiroaki*

Nuclear Engineering and Design, 229(1), p.1 - 14, 2004/04

https://doi.org/10.1016/j.nucengdes.2003.11.017

Times Cited Count：19 Percentile：75.11(Nuclear Science & Technology)

Mechanical load on cladding induced by fuel swelling in a high burn-up BWR type rod has been analyzed by a fuel performance code FEMAXI-6. The code has been developed for the analysis of LWR fuel rod behaviors in normal operation and transient conditions using FEM. During a power ramp for the high burn-up rod, instantaneous pellet swelling can significantly exceed the level that is predicted by a "steady-rate" swelling model, causing a large circumferential strain in cladding. This phenomenon has been simulated by a new swelling model to take into account the fission gas bubble growth, and as a result it has been found that the new model can give reasonable predictions on cladding diameter expansion in comparison with post-irradiation data. In addition, a pellet-clad bonding model which has been incorporated in the code to assume firm mechanical coupling between pellet outer surface and cladding inner surface has predicted the generation of bi-axial stress state in the cladding during ramp.

Journal Articles

FP plate-out and compound formation analysis in hot helium coolant of 600MWt GTHTR300 system by FP concentration difference diffusion and aerosol-condensation mechanism

Ishiyama, Shintaro

Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.106 - 119, 2004/03

https://doi.org/10.3327/taesj2002.3.106

To estimate FP deposition rate on the typical turbomachinary including gas turbin of GTHTR300 and exposure dose limit at its maintenance, FP plate-out analysis was been carried out using FP compound formation theory and FP concentration controlled diffusion mechanisum. As the results, following conclusions were derived. (1) It is found that the analysis data computed by the VICTORIA code with assumption of the correct thickness of reactive zone exhibits very good correiation with the experimental data obtained by simulated FP plate-out experiment. (2) Total dose rate and allowable work time at turbin of GTHTR300 are given as 19.2mSv/h and 2.6hours by VICTORIA code analysis.

Journal Articles

Irradiation performance of uranium-plutonium mixed nitride fuel pins in JOYO

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

JAEA Reports

Development of micro-strip gas chamber for neutron scattering; Development and simulation of encode board

Masaoka, Sei; Nakamura, Tatsuya; Yamagishi, Hideshi; Soyama, Kazuhiko

JAERI-Research 2003-012, 14 Pages, 2003/06

JAERI-Research-2003-012.pdf:1.58MB

A micro-strip gas chamber (MSGC) using helium-3 gas has been developing as a two-dimensional position sensitive neutron detector for neutron scattering experiments in a high-intensity proton accelerator facility. We have developed the encode board that have built-in FPGAs (Field-Programmable Gate Array). Logics of FPGAs can be freely programmed to carry out digital processing of signals from MSGC and information of the tracks of protons and tritons can be obtained by the encode board. Therefore, the position of neutron capture (more properly the centra of the track) can be calculated by a series of data handling system. As the result of the simulation of a neutron detection experiment, it has proved that the position resolution of less than 1.6 mm can be obtained by using this data handling system.

Journal Articles