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Konno, Chikara
JAEA-Conf 2023-001, p.143 - 146, 2024/02
I modified NJOY2016.67 to produce photonuclear ACE files which can be used in MCNP6.2 and PHITS3.27 and produced the ACE file of the JENDL-5 photonuclear sub-library. Simple test calculations with the produced ACE file supported that the produced ACE file had no serious problems.
Tada, Kenichi; Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*
Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 10 Pages, 2022/05
Nuclear data processing is an important interface between an evaluated nuclear data library and nuclear transport calculation codes. JAEA has developed a new nuclear data processing code FRENDY from 2013. FRENDY version 1 generates ACE files which are used for the continuous-energy Monte Carlo codes including PHITS, Serpent, and MCNP; it was released as an open-source software under the 2-clause BSD license in 2019. After FRENDY version 1 was released, many functions are developed: the multi-group neutron cross-section library generation, the statistical uncertainty quantification for the probability tables for unresolved resonance cross-section, the perturbation of the ACE file, and the modification of the ENDF-6 formatted nuclear data file, etc. We released FRENDY version 2 including these functions. This presentation explains the overview of FRENDY and features of the new functions implemented in FRENDY version 2.
Ono, Michitaka*; Tojo, Masayuki*; Tada, Kenichi; Yamamoto, Akio*
Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 9 Pages, 2022/05
In this paper, nuclear calculations were performed using the ACE files and the multigroup libraries created by both FRENDY and NJOY, and the impacts on the neutronics characteristics due to nuclear data processing were investigated using those libraries. MCNP was used to compare the ACE files by calculating many benchmark problems including ICSBEP and it was confirmed that the k-eff values are generally agreed with each other within the range of statistical errors. The multigroup cross sections are verified by the BWR design codes LANCR/AETNA through calculation of a commercial-grade BWR5 equilibrium core loaded with 9
9 fuels. These results indicate that fuel assembly and core characteristics are consistent with each other. From the above investigations, it was confirmed that FRENDY can provide comparable continuous/multi-group neutron cross sections with NJOY.
Yamamoto, Akio*; Tada, Kenichi; Chiba, Go*; Endo, Tomohiro*
Journal of Nuclear Science and Technology, 58(11), p.1165 - 1183, 2021/11
Times Cited Count：9 Percentile：84.69(Nuclear Science & Technology)The multi-group cross section generation capability for neutrons is implemented in the FRENDY nuclear data processing code. ACE-formatted files are used as the source of nuclear data instead of ENDF-formatted files since FRENDY already has the capability to generate pointwise cross sections in the ACE format. Verification calculations of the newly implemented capability are carried out through the comparison with the NJOY nuclear data processing code. Cross section generations for all nuclides in JENDL-4.0, -4.0u, -5
4, ENDF/B-VII.1, -VIII.0, JEFF-3.3, and TENDL-2019 are carried out without unexpected processing issue, except for Pu-238 of TENDL-2019 that includes inconsistent data. The verification results indicate that the multi-group cross sections generated by FRENDY are consistent with those generated by NJOY or the calculation results by MCNP.
Yamamoto, Akio*; Endo, Tomohiro*; Tada, Kenichi
Transactions of the American Nuclear Society, 122(1), p.714 - 717, 2020/06
A generation capability of multi-group cross sections from point-wise cross sections in ACE files is being developed as a function of the nuclear data processing code FRENDY. This presentation describes features of this function and comparison of the processing results between this function and GROUPR module in NJOY.
Tada, Kenichi
JAEA-Conf 2019-001, p.29 - 34, 2019/11
JAEA has developed a new nuclear data processing code FRENDY (FRom Evaluated Nuclear Data librarY to any application) to generate a cross-section data library from evaluated nuclear data library JENDL. In this presentation, author explains how to generate cross-section data library and overview and features of FRENDY.
Matsuda, Norihiro; Kunieda, Satoshi; Okamoto, Tsutomu*; Tada, Kenichi; Konno, Chikara
Progress in Nuclear Science and Technology (Internet), 6, p.225 - 229, 2019/01
Tada, Kenichi
Proceedings of Reactor Physics Paving the Way Towards More Efficient Systems (PHYSOR 2018) (USB Flash Drive), p.2929 - 2939, 2018/04
JAEA develops a new nuclear data processing system FRENDY. We investigated all processing methods and we focused on the probability table generation using the ladder method which is adopted in the PURR module in NJOY. To improve the probability table generation, the more sophisticated method was introduced in the calculation methods of the Chi-Squared random numbers and the complex error function. We also investigated the appropriate ladder number. To investigate the impact of the difference of the complex error function calculation method, the K
values of the benchmark experiments with the probability tables by the both methods were compared. The calculation results indicated that the appropriate ladder number is 100 and the difference of the calculation methods of the Chi-Squared random numbers and the complex error function has no significant impact on the neutronics calculation.
Tada, Kenichi; Kosako, Kazuaki*; Yokoyama, Kenji; Konno, Chikara
Nihon Genshiryoku Gakkai-Shi ATOMO
, 60(3), p.168 - 172, 2018/03
The neutronics calculation codes cannot treat the evaluated nuclear data file directly. The nuclear data processing is required to use the nuclear data file in the neutronics calculation codes. The nuclear data processing is not just a converter but also many processes to evaluate the physical values for the neutronics calculation codes. In this paper, we describe the overview of the nuclear data processing and validation of the nuclear data.
Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo
Nuclear Engineering and Design, 326, p.108 - 113, 2018/01
Times Cited Count：3 Percentile：30.05(Nuclear Science & Technology)Burn-up characteristics and criticality of impurity contained into graphite structure for commercial scale prismatic High Temperature Gas-cooled Reactor (HTGR) have been investigated. For HTGR, of which the core is filled graphite structure, the impurity contained into the graphite has unignorable poison effect for criticality. Then, GTHTR300, commercial scale HTGR, employed high grade graphite material named IG-110 to take into account the criticality effect for the reflector blocks next to fuel blocks. The fuel blocks, which should also employ IG-110, employ lower grade graphite material named IG-11 from the economic perspective. In this study, the necessity of high grade graphite material for commercial scale HTGR is reconsidered by evaluating the burn-up characteristics and criticality of the impurity. The poison effect of the impurity, which is used to be expressed by a boron equivalent, reduces exponentially like burn-up of 
B, and saturate at a level of 1 % of the initial value of boron equivalent. On the other hand, the criticality effect of the boron equivalent of 0.03 ppm, which corresponds to a level of 1 % of IG-11 shows ignorable values lower than 0.01 %
k/kk' for both of fuel blocks and reflector blocks. The impurity can be represented by natural boron without problem. Therefore, the poison effect of the impurity is evaluated with whole core burn-up calculations. As a result, it is concluded that the impurity is not problematic from the viewpoint of criticality for commercial scale HTGR because it is burned clearly until End of Cycle (EOC) even with the low grade graphite material of IG-11. According to this result, more economic electricity generation with HTGR is expected by abolishing the utilization of IG-110.
Konno, Chikara; Sato, Satoshi; Ota, Masayuki; Kwon, Saerom; Ochiai, Kentaro
Fusion Engineering and Design, 109-111(Part B), p.1649 - 1652, 2016/11
Times Cited Count：7 Percentile：55.03(Nuclear Science & Technology)Recently we have examined KERMA factors and DPA cross section data in the latest official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.0 in more detail and we found out the following new problems on the KERMA factors and DPA cross section data. (1) NJOY bugs and incorrect nuclear data generated KERMA factors and DPA cross section data of no increase with decreasing neutron energy in low neutron energy. (2) Huge helium production data caused drastically large KERMA factors and DPA cross section data in low neutron energy. (3) It seemed that NJOY could not adequately process capture cross section data in File 6, not File 12-15. (4) KERMA factors with the kinematics method are not correct for nuclear data libraries without detailed secondary particle data (energy-angular distribution data). These problems should be resolved based on our study.
Tada, Kenichi
Kaku Deta Nyusu (Internet), (113), p.7 - 23, 2016/02
This paper reports the IAEA's Consultants Meeting (CM) in Oct. 5-9, 2015. The title of the CM is "The New Evaluated Nuclear Data File Processing Capabilities".
Tada, Kenichi
Kaku Deta Nyusu (Internet), (113), p.41 - 45, 2016/02
Author prized the incentive award for nuclear data division, Atomic Energy Society of Japan in 2015. This report introduces the overview of the award-winning work.
Harada, Masahide; Watanabe, Noboru; Konno, Chikara; Meigo, Shinichiro; Ikeda, Yujiro; Niita, Koji*
Journal of Nuclear Materials, 343(1-3), p.197 - 204, 2005/06
Times Cited Count：31 Percentile：87.1(Materials Science, Multidisciplinary)For a construction of maintenance and storage scenarios for JSNS, lives of structure material need to be estimated. DPA (Displacement per Atom) was a major index of radiation damage. So we evaluated DPA value of each component. Function of the DPA calculation was equipped to the PHITS code, which was particle and heavy ion transport code. For DPA calculation, displacement cross section was necessary. Displacement cross sections of neutron below 150 MeV were processed by the NJOY code from LA150 library and those of neutron above 150MeV and proton in the all energy region were obtained from energies of fragments calculated in the PHITS. By using the PHITS, we calculated DPA values and DPA mapping. We obtained that the peak DPA values at end of 5000MWh operation were 4.1 for target vessel, 2.8 for reflector and moderator vessels, and 0.4 for proton beam windows, respectively. We estimated the target life at 1 year and the moderator life at 6 year.
Kosako, Kazuaki*; Yamano, Naoki*; Fukahori, Tokio; Shibata, Keiichi; Hasegawa, Akira
JAERI-Data/Code 2003-011, 38 Pages, 2003/07
JAERI-Data-Code-2003-011.pdf:1.29MB
The third revision of JENDL-3 (JENDL-3.3) was released in May 2002. The library is useful for many applications. For users' convenience, we have produced two JENDL-3.3 based libraries FSXLIB-J33 and MATXSLIB-J33 for transport calculation codes such as MCNP and ANISN. These two libraries are available on request.
Konno, Chikara; Ikeda, Yujiro
Journal of Nuclear Science and Technology, 39(Suppl.2), p.1037 - 1040, 2002/08
no abstracts in English
Konno, Chikara; Maekawa, Fujio; Wada, Masayuki*; Ikeda, Yujiro; Takeuchi, Hiroshi
Fusion Engineering and Design, 58-59, p.961 - 965, 2001/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Maekawa, Fujio; Sakurai, Kiyoshi; Kosako, Kazuaki*; Kume, Etsuo; Kawasaki, Nobuo*; Nomura, Yasushi; Naito, Yoshitaka*
JAERI-Data/Code 99-048, p.52 - 0, 1999/12
JAERI-Data-Code-99-048.pdf:2.34MB
no abstracts in English
Shimakawa, Satoshi
JAERI-Data/Code 99-043, p.75 - 0, 1999/09
JAERI-Data-Code-99-043.pdf:3.38MB
no abstracts in English
*; *; *; *; *; *; Harada, Hiro; ; Kume, Etsuo;
JAERI-Data/Code 97-055, 161 Pages, 1998/01
JAERI-Data-Code-97-055.pdf:4.57MB
no abstracts in English
