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Konno, Chikara
JAEA-Conf 2020-001, p.193 - 197, 2020/12
JENDL Activation Cross Section File for Nuclear Decommissioning 2017 (JENDL/AD-2017) was released in 2018. This file includes the data of neutron-induced nuclear reactions for 311 nuclides from 10
eV to 20 MeV. Thus a multi-group neutron activation cross-section library (MAXS/AD-2017) with the same format as MAXS-2015 by Dr. Okumura has been developed from JENDL/AD-2017 with PREPRO 2018 for activation calculations in nuclear facility decommissioning. MAXS/AD-2017 will be converted to ORIGEN libraries and be tested with the JPDR decommissioning data. Then MAXS/AD-2017 will be released.
Konno, Chikara; Kwon, Saerom*
Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.320 - 325, 2020/10
We found out that p-table data in the FENDL-3.1d ACE file included negative values for 33 nuclei. Thus, we studied why p-tables for heating number included negative. As a result, it was found out that partial KERMA factors became too large because the energy-balance was broken in the 33 nuclei and that FENDL-3.1d adopted kinematics KERMA factors. Then NJOY could not process adequately the 33 nuclei data, which led to negative p-tables for heating number. We prosed two solutions for this issue, produced new ACE files of FENDL-3.1d with the above two methods and confirmed that the new ACE files had no negative p-tables of the heating number.
Matsuda, Norihiro; Konno, Chikara; Ikehara, Tadashi; Okumura, Keisuke; Suyama, Kenya*
JAEA-Data/Code 2020-003, 33 Pages, 2020/03
JAEA-Data-Code-2020-003.pdf:1.85MB
Data handling modules for the radioactivity calculation code, ORIGEN-S, are developed for the reliable evaluations of radioactivity inventory. By using these modules, an activation cross-section data library for the ORIGEN-S code is updated easily and effectively based on a facility-specific neutron spectrum and multi-group neutron activation cross-section library for decommissioning of nuclear facilities, MAXS2015. In order to guarantee the reliability of the radioactivity calculations, functions of data verification in a visual way and numerical comparison between before and after the data processing are also prepared.
Kwon, Saerom*; Konno, Chikara; Ota, Masayuki*; Kasugai, Atsushi*
Journal of Nuclear Science and Technology, 57(3), p.344 - 351, 2020/03
Times Cited Count：0 Percentile：100(Nuclear Science & Technology)We found out that there was a questionable iron DPA value just above 20 MeV neutron energy in neutronics analyses of A-FNS using FENDL-3.1d. Our detailed investigation on the iron data in FENDL-3.1d figured out that residual nucleus production yield data of 
Fe just above 20 MeV had a problem, which caused a sharp spike just above 20 MeV in the DPA cross section of Fe. Thus we modified the yield data of Fe and verified that the questionable DPA value disappeared using the modified data. We also examined DPA cross sections of other nuclei in FENDL-3.1d. It was found out that DPA cross sections of more than 70% of nuclei in FENDL-3.1d have similar problems as that of Fe.
Trkov, A.*; Griffin, P. J.*; Simakov, S. P.*; Greenwood, L. R.*; Zolotarev, K. I.*; Capote, R.*; Aldama, D. L.*; Chechev, V.*; Destouches, C.*; Kahler, A. C.*; et al.
Nuclear Data Sheets, 163, p.1 - 108, 2020/01
Times Cited Count：8 Percentile：2.11(Physics, Nuclear)The version II of the International Reactor Dosimetry and Fusion File (IRDFF-II) has been released as a consistent set of nuclear data for fission and fusion neutron metrology applications up to 60 MeV neutron energy. The library is intended to support: (a) applications in research reactors; (b) safety and regulatory applications in the nuclear power generation in commercial fission reactors; and c) material damage studies in support of the research and development of advanced fusion concepts. The paper describes the contents of the library, documents the thorough verification process used in its preparation, and provides an extensive set of validation data gathered from a wide range of neutron benchmark fields. The new library is expected to become the international reference in neutron metrology for multiple applications.
Konno, Chikara; Kwon, Saerom*
JAEA-Conf 2019-001, p.167 - 172, 2019/11
TENDL (TALYS-based Evaluated Nuclear Data Library) has been used as a standard nuclear data library worldwide, particularly in Europe. Since 2016 we also have used the official ACE files of TENDL-2015 for our study, where we found two problems. (1) There are no probability table data in the neutron sub-library ACE files of most of the nuclei with unresolved resonance data. Calculated results are not correct in the case that the self-shielding effect in the unresolved resonance region is large. (2) There are no secondary gamma data in a lot of the ACE files not only of the neutron sub-library but also of the proton, deuteron, triton, and helium sub-libraries. This is due to an inadequate NJOY input. MCNP reads particle production data in the official ACE files as gamma production data incorrectly and produces wrong secondary gammas. It is noted that the official ACE files in the latest TENDL-2017 still have these problems except for those of main nuclei, which have been revised based on our study.
Kwon, Saerom*; Konno, Chikara; Ota, Masayuki*; Ochiai, Kentaro*; Sato, Satoshi*; Kasugai, Atsushi*
Fusion Engineering and Design, 144, p.209 - 214, 2019/07
Times Cited Count：0 Percentile：100(Nuclear Science & Technology)We performed a TENDL-2017 benchmark test with iron shielding experiments by using 40 and 65 MeV neutrons, in order to verify a nuclear data library above 20 MeV for neutronics analyses of A-FNS. We found out that the calculated neutron spectra with TENDL-2017 unnaturally increased near 30 MeV. We figured out that incorrect secondary neutron spectrum data in 
Fe, Fe and 
Fe at 30 MeV caused the increase of the neutron flux. Similar problems occurred in a lot of nuclei of TENDL-2017, TENDL-2015 and FENDL-3.1d from TENDL-2010 and TENDL-2011.
Konno, Chikara
Progress in Nuclear Science and Technology (Internet), 6, p.117 - 121, 2019/01
SCALE6.2.1 was released in 2016 and has been used worldwide. It includes new AMPX format files (AMPX MG libraries) of ENDF/B-VII.0 and ENDF/B-VII.1 and a new nuclear data processing code AMPX-6, which produces AMPX MG libraries. Thus we produce an AMPX MG library of JENDL-4.0 in order to disseminate JENDL-4.0. Neutron and 
spectra inside an iron or other material sphere of 1 m in radius with a 20 MeV neutron source at the center were calculated with a one-dimensional Sn code ANISN for testing the JENDL-4.0 AMPX MG library. As a result, it was verified that the JENDL-4.0 AMPX MG library had no problems. Note that the self-shielding correction for AMPX MG libraries was still inadequate in shielding calculations.
Matsuda, Norihiro; Kunieda, Satoshi; Okamoto, Tsutomu*; Tada, Kenichi; Konno, Chikara
Progress in Nuclear Science and Technology (Internet), 6, p.225 - 229, 2019/01
Konno, Chikara; Kwon, Saerom*; Fischer, G.*
ANS RPSD 2018; 20th Topical Meeting of the Radiation Protection and Shielding Division of ANS (CD-ROM), 4 Pages, 2018/08
IAEA released two patches for TRANSX2.15 for the MATXS file of FENDL-2.0 in 1998. The first patch is required for all MATXS files, but it is not known well because it is not officially included to TRANSX2.15. Recently we investigated effects of the patch with a simple calculation. As a result, it is found out that the patch solves an overestimation problem of neutron fluxes in Sn calculations with self-shielding corrected multigroup libraries generated with the original TRANSX code. This patch should be officially included to TRANSX2.15 because it is essential.
Sato, Satoshi*; Konno, Chikara; Nakashima, Hiroshi; Shionaga, Ryosuke*; Nose, Hiroyuki*; Ito, Yuji*; Hashimoto, Hirohide*
Journal of Nuclear Science and Technology, 55(4), p.410 - 417, 2018/04
Times Cited Count：1 Percentile：77.66(Nuclear Science & Technology)In order to enhance the neutron shielding performance, we developed concrete with boron of more than 10 wt%. We performed a neutron shielding experiment using the mockup of the newly developed boron-loaded concrete and DT neutrons at FNS in JAEA, and measured the reaction rates of the 
Nb(n,2n)
Nb and 
Au(n,)
Au reactions in the mockup. The calculations were conducted by using MCNP-5.14 and FENDL-2.1. The calculation results agreed well with the measured ones, and we confirmed that the accuracy was very good on the atomic composition data of the boron-loaded concrete and their nuclear data. In addition, we calculated effective dose rates and reaction rates of the 
Co(n,)
Co and 
Eu(n,)
Eu reactions in the boron-loaded concrete and other concretes. It is concluded that the boron-loaded concrete has much better shielding performance for DT neutrons than other concretes.
Tada, Kenichi; Kosako, Kazuaki*; Yokoyama, Kenji; Konno, Chikara
Nippon Genshiryoku Gakkai-Shi, 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.
Konno, Chikara; Kwon, Saerom*; Ota, Masayuki*; Sato, Satoshi*
JAEA-Conf 2017-001, p.117 - 122, 2018/01
The revised version of FENDL-3, FENDL-3.1b was released in October, 2015. Thus we have tested FENDL-3.1b neutron sub-library for the problems we reported to IAEA before. Most of the MATXS files above 20 MeV had no scattering matrix data of non-elastic scattering, but this problem was fixed by re-processing FENDL-3 with NJOY2012.50. As for the problem on KERMA factors and DPA data, IAEA revised the wrong Q value of the capture reaction in 
N and re-calculated KERMA factors and DPA data with NJOY2012.50. It was confirmed that most of the KERMA factors and DPA data were revised correctly except for huge gas production cross-section data. However a new problem on NJOY processing of gas production data was found out. It was pointed out that this problem was due to a bug of NJOY. Additionally we investigated a trouble on 
Sn and 
Sn NJOY processing at IAEA and specified that one of NJOY patches caused this trouble.
2 benchmark test with shielding experiments at QST/TIARAKwon, Saerom*; Konno, Chikara; Ota, Masayuki*; Sato, Satoshi*; Ochiai, Kentaro*
JAEA-Conf 2017-001, p.123 - 128, 2018/01
The -version of ENDF/B-VIII, ENDF/B-VIII2, was released in August, 2016. Thus we studied whether the overestimation problems due to the 
O and Fe data of ENDF/B-VII.1 were corrected in the iron and concrete shielding experiments with 40 and 65 MeV neutrons at TIARA. We produced the ACE files of ENDF/B-VIII2 with the NJOY2012.50 code and used the MCNP-5 code for this analysis. The nuclear data libraries, ENDF/B-VII.1, FENDL-3.1b and JENDL-4.0/HE, were also used for comparison. The following results were obtained; (1) the drastic overestimation of around 40 MeV due to the 5Fe data was improved, (2) the overestimation for around 65 MeV due to the Fe data was also slightly improved, though it was worse than that with FENDL-3.1b, (3) the drastic overestimation due to the O data was not improved. The final version of ENDF/B-VIII should also be modified based on these results.
Kwon, Saerom*; Ota, Masayuki*; Sato, Satoshi*; Konno, Chikara; Ochiai, Kentaro*
Fusion Engineering and Design, 124, p.1161 - 1164, 2017/11
Times Cited Count：1 Percentile：81.79(Nuclear Science & Technology)Copper is used as a material for superconducting coil in magnetic confinement fusion reactor and for accelerator-driven neutron source such as IFMIF. In our previous copper benchmark experiment, we had pointed out that the elastic scattering and capture reaction data of the copper had included some problems in the resonance region, which had caused a large underestimation of reaction rates of non-threshold reactions. In order to corroborate this issue, we carried out a new benchmark experiment on copper with graphite in the neutron field with more low energy neutrons. We measured reaction rates using the activation foils. We analyzed the experiment with MCNP code and the latest nuclear data libraries. As a result, the calculated reaction rates related to low energy neutrons, still excessively underestimated the measured ones as in the previous benchmark experiment. We also tested the nuclear data of copper modified in the previous study, where the elastic scattering and capture reaction cross section of copper. Then the calculated reaction rates with the modified copper nuclear data reproduced the measured ones well. It was revealed that the modification of the specific cross sections had been sufficient in the neutron field with more low energy neutrons.
Kwon, Saerom*; Ota, Masayuki*; Sato, Satoshi*; Konno, Chikara; Ochiai, Kentaro*
Fusion Science and Technology, 72(3), p.362 - 367, 2017/10
Times Cited Count：1 Percentile：81.79(Nuclear Science & Technology)Lead is a candidate material as a neutron multiplier, a tritium breeder and a coolant in nuclear fusion reactor system, and a ray shielding for beam dump or shielding of components in accelerator-driven neutron source such as IFMIF. A benchmark experiment on lead with DT neutrons had been performed at JAEA/FNS seven, where the reaction rates related to neutrons below a few keV had included background neutrons scattered in concrete walls of the experiment room. Thus, we designed and carried out a new benchmark experiment with a lead assembly covered with Li
O blocks absorbing background neutrons. We successfully measured reaction rates of the non-threshold reactions with the activation foil method. The experiment was analyzed with MCNP code and the latest nuclear data libraries. All the calculated reaction rates (C) tended to underestimate the experimental ones (E) with the depth of the lead assembly. Although reasons of the underestimation have not been specified yet, we discovered that there are remarkable different tendencies of C/Es each reaction rate among the nuclear data libraries.
Konno, Chikara; Tada, Kenichi; Kwon, Saerom*; Ota, Masayuki*; Sato, Satoshi*
EPJ Web of Conferences, 146, p.02040_1 - 02040_4, 2017/09
Times Cited Count：0 Percentile：100So far we pointed out that KERMA factors and DPA cross-section data of a lot of nuclei in the official ACE file were different among nuclear data libraries for the following reasons; (1) incorrect nuclear data, (2) NJOY bugs, (3) huge helium production cross section data, (4) mf6 mt102 data, (5) no secondary particle data (energy-angular distribution data). Now we compare the KERMA factors and DPA cross section data included in the official ACE files of JENDL-4.0, ENDF/B-VII.1 and JEFF-3.2 in more detail. As a result, we find out new reasons of differences among the KERMA factors and DPA cross section data in the three nuclear data libraries. The reasons are categorized to no secondary charged particle data, no secondary data, wrong secondary spectra, wrong production yields and mf12-15 mt3 data for the capture reaction, some of which seem to be unsupported with NJOY. The ACE files of JENDL-4.0, ENDF/B-VII.1 and JEFF-3.2 with these problems should be revised based on this study.
Konno, Chikara; Matsuda, Norihiro; Kwon, Saerom*; Ota, Masayuki*; Sato, Satoshi*
EPJ Web of Conferences, 153, p.01024_1 - 01024_6, 2017/09
Times Cited Count：2 Percentile：11.38As a benchmark test of JENDL-4.0/HE released in 2015, we have analyzed concrete and iron shielding experiments with the 40 and 65 MeV neutron sources at TIARA in JAEA by using MCNP5 and ACE files processed from JENDL-4.0/HE with NJOY2012. As a result, it was found out that the calculation results with JENDL-4.0/HE agreed with the measured ones in the concrete experiment well, while they underestimated the measured ones in the iron experiment more for the thicker assemblies. We examined JENDL-4.0/HE in detail and it was considered that the larger non-elastic scattering cross sections of Fe caused the underestimation in the calculation with JENDL-4.0/HE for the iron experiment.
Ota, Masayuki*; Kwon, Saerom*; Sato, Satoshi*; Konno, Chikara; Ochiai, Kentaro*
Fusion Engineering and Design, 114, p.127 - 130, 2017/01
Times Cited Count：0 Percentile：100(Nuclear Science & Technology)A new fusion neutron source is now under consideration in Japan. Type 316L stainless steel (SUS316L) which is a structural material of the target-system contains a few percent of molybdenum. In our previous benchmark experiment on molybdenum at JAEA/FNS, we found problems of the cross section data above a few hundred eV in Mo. We perform a new benchmark experiment on Mo with graphite in order to validate the Mo data in the lower energy region. Several dosimetry reaction rates and fission rates are measured in the assembly and compared with the calculated values with the Monte-Carlo transport code MCNP5-1.40 and the recent nuclear data libraries. It is suggested that the (n,) cross section of 
Mo is underestimated in the tail region below the large resonance at 45 eV in the recent nuclear data libraries.
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：1 Percentile：84.04(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.
