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Yokoyama, Kenji; Sugino, Kazuteru; Ishikawa, Makoto; Maruyama, Shuhei; Nagaya, Yasunobu; Numata, Kazuyuki*; Jin, Tomoyuki*
JAEA-Research 2018-011, 556 Pages, 2019/03
JAEA-Research-2018-011.pdf:19.53MB
JAEA-Research-2018-011-appendix1(DVD-ROM).zip:433.07MBJAEA-Research-2018-011-appendix2(DVD-ROM).zip:580.12MBJAEA-Research-2018-011-appendix3(DVD-ROM).zip:9.17MB
We have developed a new unified cross-section set ADJ2017, which is an improved version of the unified cross-section set ADJ2010 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses; the values are stored in the standard database for FBR core design via the cross-section adjustment methodology, which integrates with the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. The ADJ2017 is based on Japan's latest nuclear data library JENDL-4.0 as in the previous version of ADJ2010, and it incorporates more information on integral experimental data sets related to minor actinides (MAs) and degraded plutonium (Pu). In the creation of ADJ2010, a total of 643 integral experimental data sets were analyzed and evaluated, and 488 of the integral experimental data sets were finally selected to be used for the cross-section adjustment. In contrast, we have evaluated a total of 719 data sets, and eventually adopted 620 integral experimental data sets to create ADJ2017. ADJ2017 shows almost the same performance as ADJ2010 for the main neutronic characteristics of conventional sodium-cooled MOX-fuel fast reactors. In addition, for the neutronic characteristics related to MA and degraded Pu, ADJ2017 improves the C/E values of the integral experimental data sets, and reduces the uncertainty induced by the nuclear data. ADJ2017 is expected to be widely used in the analysis and design research of fast reactors. Moreover, it is expected that the integral experimental data sets used for ADJ2017 can be utilized as a standard database of FBR core design.
Takase, Kazuyuki; Akimoto, Hajime
Nuclear Fusion, 41(12), p.1873 - 1883, 2001/12
Times Cited Count：6 Percentile：21.41(Physics, Fluids & Plasmas)no abstracts in English
Takase, Kazuyuki; Akimoto, Hajime
Proceedings of IAEA 18th Fusion Energy Conference (CD-ROM), 5 Pages, 2001/00
no abstracts in English
Takase, Kazuyuki
Fusion Engineering and Design, 51-52(Part.B), p.631 - 639, 2000/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
*; *
JNC TJ7400 2000-007, 222 Pages, 2000/03
JNC-TJ7400-2000-007.pdf:13.09MB
no abstracts in English
*; Kitada, Takanori*; Tagawa, Akihiro; *; Takeda, Toshikazu*
JNC TJ9400 2000-006, 272 Pages, 2000/02
JNC-TJ9400-2000-006.pdf:9.69MB
Investigation was made on the follwing three themes as a part of the improvement of reactor physics analysis method for FBR with various core concept. Part 1: Investigation of Error Estimation of Neutron Spectra in FBR and Suggestions to Improve the Accuracy. In order to improve the spectrum unfolding method used in fast experimental reactor JOYO, a trial was made to evaluate the error in the estimated neutron spectrum, cause by cause. And the evaluated errors were summed up to obtain the most probable and reasonable error as possible. The summed up error was found relatively small compared to the error caused by the uncertainty of cross section data: most of the error in the spectrum unfolding method can be attributed to the error in cross sections. It was also found that the error due to the fission spectrum causes a considerable error in the high energy neutron spectrum which is over several MeV. Part 2: Study on Reactor Physics Analysis Method for Gas-Cooled FBR. In gas-cooled FBR, the portion of coolant channels in core volume is larger than sodium-cooled FBR. This leads to strong neutron streaming effects. For sodium-cooled FBR, several methods were proposed to evaluate the neutron streaming effect, however, these methods can not be used directly to gas-cooled reactor because the direction dependent diffusion coefficient becomes infinitive along the direction pararel to the coolant chammel. In this study, a new method is proposed to evaluate the neutron streaming effect, based on the method taking the axial buckling into consideration, which method was originally proposed by K
hler. Part 3: Study on Reactor Physics Analysis Method for Water-Cooled FBR An investigation was made on low-moderated water-cooled FBR, on the point that the ordinary used analysis method for FBR may give considerable difference in results in such core. In light water reactors, it is well known that the space dependence of self-shielding effect of heavy nuclides are considerably ...
; Kawai, Masashi; ; ; ; Okubo, Toshiyuki;
PNC TN9410 94-094, 69 Pages, 1994/03
As the final step of development of a control rod guide system (rod guider), it has been conducted to display information concerning the rod guider operation and the current plant conditions on a CRT in Japanese, to tell information for operators and to predict more accurately the stroke of each control rod to be driven at a reactor power adjustment operation since 1992, in oder to improve a function of man-machine interface of the rod guider. The development of the rod guider was completed with the following good results as expected. (1)The function of man-machine interface of the rod guider was improved to display information on a CRT in Japanese. (2)The voice guidance contributed to obtain higher operational reliability and to lower the load on the operators. (3)The rod guider provided the same function as operation manuals of Joyo because all operation procedures in the manuals were inputted into the rod guider. And the timely operation guides provided by the rod guider evabled non-skilled operators to control the reactor in the same manner as the experienced ones. (4)The difference of the predicted and actual rod stroke to be driven at the reactor power adjustment operation was reduced to less than 
0.2mm by both the existing inverse multiplication method and the experiences gained in the past five reactor power adjustment operations.
Nakakawa, Masayuki; Mori, Takamasa; Kosako, Kazuaki*; Oyama, Yukio; Nakamura, Tomoo
JAERI-M 89-154, 178 Pages, 1989/10
no abstracts in English
Ishikawa, Makoto
no journal, ,
As a topic of Nuclear-data Tutorial, the author will make a lecture on the method to use nuclear-data covariance, which includes the background of nuclear-data covariance use, the physical meaning of covariance, the method to evaluate the prediction accuracy of nuclear core-parameters with covariance, and how to improve the nuclear design accuracy with covariance.
