Kwon, Saerom*; Konno, Chikara; Ota, Masayuki*; Ochiai, Kentaro*; Sato, Satoshi*; Kasugai, Atsushi*
Fusion Engineering and Design, 144, p.209 - 214, 2019/07
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.
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.
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
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.
Kwon, 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
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
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 LiO 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 (Internet), 146, p.02040_1 - 02040_4, 2017/09
So 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 (Internet), 153, p.01024_1 - 01024_6, 2017/09
As 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
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
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.
Iwamoto, Yosuke; Konno, Chikara
Kaku Deta Nyusu (Internet), (115), p.24 - 32, 2016/10
The technical meeting of nuclear reaction data and uncertainties for radiation damage was held at IAEA Headquarters in June 2016. This Meeting was organized to implement the recommendation of the IAEA Coordinated Research Project (CRP) "Primary Radiation Damage Cross Sections" to analyze the accuracy and consistency of the radiation damage-relevant nuclear data in the major nuclear data evaluations with the eventual goal of identifying the most reliable data and providing quantitative uncertainty estimates. Participants have considered the status of the primary nuclear data, such as reaction recoils spectra in the latest releases of nuclear data libraries, and the ways of deriving the damage quantities KERMA, dpa cross sections and gas production cross sections as well as the recipes for an assessment of their uncertainties. This report contains the contemporary view of the Meeting participants on these issues in the form of a consolidated set of statements and recommendations.
Sato, Satoshi*; Kwon, Saerom*; Ota, Masayuki*; Ochiai, Kentaro*; Konno, Chikara
Proceedings of 26th IAEA Fusion Energy Conference (FEC 2016) (CD-ROM), 8 Pages, 2016/10
In the integral experiments on tungsten, vanadium and copper performed with the DT neutron source at JAEA/FNS over 20 years ago, the calculated results had largely underestimated the measured ones sensitive to low energy neutrons in the experiments. Since background neutrons scattered in the concrete wall of the experimental room were considered to cause these underestimations, in this study we performed new integral experiments with these materials covered with LiO blocks absorbing background neutrons. We also performed similar integral experiments on molybdenum and titanium. We analyzed these experiments by using MCNP5-1.40 with ENDF/B-VII.1, JEFF-3.2 and JENDL-4.0. The large underestimations observed in the previous tungsten and vanadium experiments disappeared in the present experiments, which led that the nuclear data of tungsten and vanadium had no problem. On the other hand, the underestimation was not improved so much in the copper experiment, and the calculation results also did not show good agreements with the measured ones in the molybdenum and titanium experiments. Detailed analyses with partly modified nuclear data clarified the problems of the nuclear data libraries on copper, molybdenum and titanium.
Konno, Chikara; Kwon, Saerom; Ota, Masayuki; Sato, Satoshi
JAEA-Conf 2016-004, p.233 - 238, 2016/09
We compared the KERMA factors and DPA cross section data included in the official ACE and MATXS files of JENDL-4.0 with those of ENDF/B-VII.1 and JEFF-3.2. As a result, they were different from those of ENDF/B-VII.1 and JEFF-3.2 in a lot of nuclei, which was considered to be caused by the following new problems; (1) NJOY bugs, (2) huge helium production cross section data, (3) production data format in the nuclear data, (4) no detailed secondary particle data (energy-angular distribution data). The ACE and MATXS files of JENDL-4.0 with these problems should be revised based on this study.
Konno, Chikara; Kwon, Saerom; Ota, Masayuki; Sato, Satoshi
JAEA-Conf 2016-004, p.239 - 242, 2016/09
In order to specify reasons of the discrepancy between the calculated and measured results in analyses of benchmark experiments, some parts of some isotope data in nuclear data files are often modified and the modifies nuclear data files are processed with the NJOY code and the new ACE or MATXS files are used. However it is not easy to modify capture and elastic scattering data below 1 MeV with resonance data. Thus we devised a simple method to use capture and elastic scattering cross section data generated from resonance data with the NJOY code. This method was applied to detailed analyses of copper and molybdenum benchmark experiments at JAEA/FNS and it was demonstrated that this method was a very powerful tool.
Purazuma, Kaku Yugo Gakkai-Shi, 92(4), p.261 - 265, 2016/04
Useful information to not only beginners but also experts is introduced mainly for important points, which are basic but surprisingly unknown in, about calculation codes and nuclear data libraries in nuclear analyses for fusion reactors.
Konno, Chikara; Ota, Masayuki; Kwon, Saerom; Ochiai, Kentaro; Sato, Satoshi
JAEA-Conf 2015-003, p.131 - 136, 2016/03
We carried out the benchmark tests of the general-purpose data library for neutron-induced reactions in FENDL-3.0 with the integral experiments at JAEA/FNS, JAEA/TIARA and Osaka Univ./OKTAVIAN. We also tested the MATXS files of FENDL-3.0 with a simple calculation model and compared KERMA and DPA data included in the ACE and MATXS files of FENDL-3.0 with those in other nuclear data libraries. In this symposium we present the following problems in FENDL-3.0 found out in our study; (1) The O data above 20 MeV in FENDL-3.0 should be revised. (2) The most MATXS files in FENDL-3.0 have no energy-angular distribution data for the non-elastic scattering reaction. (3) Some of KERMA and DPA data included in the ACE and MATXS files of FENDL-3.0 should be revised.