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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.

Journal Articles

Development of experimental and analytical technologies for fission product chemistry under LWR severe accident condition

Miyahara, Naoya; Miwa, Shuhei; Nakajima, Kunihisa; Osaka, Masahiko

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 9 Pages, 2017/09

This paper presents the development of a reproductive experimental setup for FP release and transport and an analysis tool considering chemical reaction kinetics for the construction of the FP chemistry database. The performance test of the reproductive experimental setup TeRRa using CsI compounds show that TeRRa can reproduce well a FP chemistry-related behavior such as aerosol formation, growth and deposition behavior. An analytical tool has been developed based on the commercial ANSYS-FLUENT code. Some additional models was added to evaluate detailed FP chemistry during release and transport in this study. A test analysis simulating the CsI heating test in steam atmosphere was carried out to demonstrate the performance of the improved code. The result shows the appropriateness of the additional models.

Journal Articles

Release of radioactive materials from high active liquid waste in small-scale hot test for boiling accident in reprocessing plant

Yamane, Yuichi; Amano, Yuki; Tashiro, Shinsuke; Abe, Hitoshi; Uchiyama, Gunzo; Yoshida, Kazuo; Ishikawa, Jun

Journal of Nuclear Science and Technology, 53(6), p.783 - 789, 2016/06

 Times Cited Count:1 Percentile:76.09(Nuclear Science & Technology)

The release behavior of radioactive materials from high active liquid waste (HALW) has been experimentally investigated under boiling accident conditions. In the experiments using HALW obtained through laboratory scale reprocessing, release ratio was measured for the FP nuclides such as Ru, $$^{99}$$Tc, Cs, Sr, Nd, Y, Mo, Rh and actinides such as $$^{242}$$Cm, $$^{241}$$Am. As a result, the release ratio was 0.20 for Ru and 1$$times$$$$10^{-4}$$ for the FP and Ac nuclides. Ru was released into the gas phase in the form of both mist and gas. For its released amount, weak dependency was found to the initial concentration in the test solution. The release ratio decreased with the initial concentration. For other FP nuclides and actinides as non-volatile, released into the gas phase in the form of mist, the released amount increased with the initial concentration. The release ratio of Ru and NOx concentration increased with temperature of the test solutions. They were released almost at the same temperature between 200 and 300$$^{circ}$$C. Size distribution of the mist and other particle was measured.

JAEA Reports

Boron release kinetics from mixed melts of boron carbide, stainless steel and Zircaloy; A Literature review on the behavior of control rod materials under severe accidents

Di Lemma, F. G.; Miwa, Shuhei; Osaka, Masahiko

JAEA-Review 2016-007, 27 Pages, 2016/03

JAEA-Review-2016-007.pdf:1.88MB

During a nuclear power plant Severe Accident, complex boron melts can be formed, due to interaction of the control rods with the cladding materials. These can affect ultimately the source term assessment. This review will describe the results of previous studies on boron carbide/stainless steel/Zircaloy (B$$_{4}$$C/SS/Zry) melts, which will finally provide guidance for the needs of future experiments. This review showed that models for the behavior of complex B$$_{4}$$C/SS/Zry melts are limited, and unsuccessful in simulating core degradation, thus the improvement of the database for B$$_{4}$$C/SS/Zry melts is needed. Our experimental plan aims in providing thermodynamics and kinetics models for such melts, with the final aim of improving boron modelling in SA codes analysis and of understanding its effect on fission products behavior.

JAEA Reports

Application of FORNAX-A

Aihara, Jun; Ueta, Shohei; Nishihara, Tetsuo

JAEA-Technology 2015-040, 32 Pages, 2016/02

JAEA-Technology-2015-040.pdf:0.83MB

Original FORNAX-A is a calculation code for amount of fission product (FP) released from fuel rods of pin-in-type high temperature gas-cooled reactors (HTGRs). This report is for explanation what calculations become possible with minor changed FORNAX-A.

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

Prediction of fuel performance and fission gas release behavior during normal operation of the High Temperature Engineering Test Reactor by JAERI and FZJ modeling approach

Sawa, Kazuhiro; Ueta, Shohei; Sumita, Junya; Verfondern, K.*

Journal of Nuclear Science and Technology, 38(6), p.411 - 419, 2001/06

no abstracts in English

Journal Articles

Quick look of first VEGA test and fabrication study of thoria components

Nakamura, Takehiko; Hidaka, Akihide; Kudo, Tamotsu; Hayashida, Retsu*; Otomo, Takashi; Nakamura, Jinichi; Uetsuka, Hiroshi

JAERI-Conf 2000-015, p.201 - 209, 2000/11

no abstracts in English

JAEA Reports

Research program (VEGA) on the fission product release from irradiated fuel

Nakamura, Takehiko; Hidaka, Akihide; Hashimoto, Kazuichiro; Harada, Yuhei; Nishino, Yasuharu; Kanazawa, Hiroyuki; Uetsuka, Hiroshi; Sugimoto, Jun

JAERI-Tech 99-036, 34 Pages, 1999/03

JAERI-Tech-99-036.pdf:1.55MB

no abstracts in English

JAEA Reports

Metallic fission product release from failed coated fuel particles; ICF-51H capsule irradiation test

Tobita, Tsutomu; Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kosaku; Sekino, Hajime; Iida, Shozo;

JAERI-Research 96-014, 34 Pages, 1996/03

JAERI-Research-96-014.pdf:2.15MB

no abstracts in English

JAEA Reports

CORCON-Mod3 analysis of SURC experiments on molten core concrete interaction

J.Yan*; Maruyama, Yu; Sugimoto, Jun

JAERI-Tech 95-052, 27 Pages, 1995/12

JAERI-Tech-95-052.pdf:1.32MB

no abstracts in English

JAEA Reports

A Study of silver behavior in gas-turbine high temperature gas-cooled reactor

Sawa, Kazuhiro; Tanaka, Toshiyuki

JAERI-Research 95-071, 23 Pages, 1995/11

JAERI-Research-95-071.pdf:0.53MB

no abstracts in English

Journal Articles

Release of fission products from plate type fuel at elevated temperature

; ; Iwai, Takashi; Saito, Junichi; Nakagawa, Tetsuya; Oyamada, Rokuro; Saito, Minoru

JAERI-M 93-016, 0, p.259 - 267, 1993/02

no abstracts in English

JAEA Reports

Plate-out distribution of iodine in a high temperature gas cooling in-pile loop facility

*; Endo, Yasuichi; ; Itabashi, Yukio; ; Yokouchi, Iichiro; Ando, Hiroei

JAERI-M 92-212, 62 Pages, 1993/01

JAERI-M-92-212.pdf:2.09MB

no abstracts in English

Journal Articles

Release behavior of metallic fission products from HTGR fuel particles at 1600 to 1900$$^{circ}$$C

Minato, Kazuo; Ogawa, Toru; Fukuda, Kosaku; Sekino, Hajime; ; ;

Journal of Nuclear Materials, 202, p.47 - 53, 1993/00

 Times Cited Count:46 Percentile:3.65

no abstracts in English

Journal Articles

Verification of fission product release model from High Temperature Engineering Test Reactor fuel

Sawa, Kazuhiro; Shiozawa, Shusaku; Fukuda, Kosaku;

Journal of Nuclear Science and Technology, 29(9), p.842 - 850, 1992/09

no abstracts in English

Journal Articles

Release of fission products from silicide fuel at elevated temperatures

; Saito, Minoru; Oyamada, Rokuro; ; ; Saito, Junichi; Iwai, Takashi; ; Nakagawa, Tetsuya

Nucl. Saf., 33(3), p.334 - 343, 1992/07

no abstracts in English

JAEA Reports

Evaluation of fission product sources for shielding design of HTTR

Sawa, Kazuhiro; Murata, Isao; ; Shiozawa, Shusaku

JAERI-M 91-198, 58 Pages, 1991/11

JAERI-M-91-198.pdf:1.58MB

no abstracts in English

JAEA Reports

Siting evaluation of the High Temperature Engineering Test Reactor

Sawa, Kazuhiro; Shiozawa, Shusaku; Shindo, Masami; Tazawa, Yujiro*; *; *; *;

JAERI-M 91-158, 69 Pages, 1991/10

JAERI-M-91-158.pdf:1.85MB

no abstracts in English

Journal Articles

Release behavior of metallic fission products from pyrocarbon-coated uranium-dioxide particles at extremely high temperatures

; Fukuda, Kosaku

Journal of Nuclear Science and Technology, 27(4), p.320 - 332, 1990/04

no abstracts in English

36 (Records 1-20 displayed on this page)