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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:4 Percentile:37.41(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.
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
Times Cited Count:3 Percentile:26.88(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:6 Percentile:48.39(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 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.
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:0.00(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:8 Percentile:57.96(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.
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.
Sakasai, Kaoru; To, Kentaro; Nakamura, Tatsuya; Ochiai, Kentaro; Konno, Chikara
Proceedings of 2014 IEEE Nuclear Science Symposium and Medical Imaging Conference; 21st International Symposium on Room-Temperature Semiconductor X-ray and -ray detectors (NSS/MIC 2014), Vol.3 , p.1834 - 1839, 2016/05
no abstracts in English
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.
Konno, Chikara; Ota, Masayuki; Kwon, Saerom; Ochiai, Kentaro; Sato, Satoshi
JAEA-Conf 2015-003, p.125 - 130, 2016/03
At the last nuclear data symposium we presented the detailed analyses of the iron and concrete shielding experiments with 40 and 65 MeV neutrons at TIARA in JAEA in order to validate FENDL-3.0 and JENDL/HE-2007 and pointed out that calculation results with JENDL/HE-2007 underestimated the measured neutron spectra and calculated ones with FENDL-3.0 in the iron experiment with 65 MeV neutrons. Thus we studied reasons of this underestimation in detail. As a result, we specified that the larger non-elastic scattering cross section data of Fe in JENDL/HE-2007 caused the underestimation. The non-elastic scattering data of Fe in JENDL/HE-2007 should be revised.
Konno, Chikara; Ota, Masayuki; Kwon, Saerom; Ochiai, Kentaro; Sato, Satoshi
JAEA-Conf 2015-003, p.137 - 141, 2016/03
For fusion reactor nuclear analyses we produce new nuclear group constant sets, FUSION-F21.175 (neutron: 175 groups, : 42 groups, P5 approximation) and FUSION-F21.42 (neutron: 42 groups, : 21 groups, P5 approximation), similar with FUSION-J3 and FUSION-40 from FENDL-2.1 with the TRANSX code. The materials in these sets are H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-12, N-14, O-16, F-19, Na-23, Mg, Al-27, Si, P-31, S, K, Ca, Ti, V-51, Cr, Mn-55, Fe, Co, Ni, Cu, Zr, Nb-93, Mo, Cd, W, Pb, Bi-209, Cl, Ta-181, Sn and Ga. It should be noted that the self-shielding effect is not corrected in these libraries. KERMA, DPA and gas production libraries are also prepared from the MATXS files with TRANSX. Several test calculations are carried out in order to validate these nuclear group constant sets. They suggest that these group constant sets have no problem.
Fujita, Hiroe*; Yuyama, Kenta*; Li, X.*; Hatano, Yuji*; Toyama, Takeshi*; Ota, Masayuki; Ochiai, Kentaro; Yoshida, Naoaki*; Chikada, Takumi*; Oya, Yasuhisa*
Physica Scripta, 2016(T167), p.014068_1 - 014068_5, 2016/02
Times Cited Count:33 Percentile:82.96(Physics, Multidisciplinary)The irradiation defects were introduced by Fe irradiation, fission neutron irradiation and D-T neutron irradiation. After the irradiation, the deuterium ions (D) implantation was performed and the D retention behavior was evaluated by thermal desorption spectroscopy. The experimental results indicated that dense vacancies and voids within the shallow region near the surface were introduced by Fe irradiation. The trapping state of D by vacancies and void were clearly controlled by the damage concentration and the voids would become the most stable D trapping site. For fission neutron irradiated W, most of the D was adsorbed on the surface and/or trapped by dislocation loops and no vacancies and voids for D trapping due to its lower damage concentration. D trapping by vacancies were found in the bulk of D-T neutron irradiated W, indicating that the neutron energy distribution could make a large impact on irradiation defect formation and the D retention behavior.
Konno, Chikara; Ochiai, Kentaro; Sato, Satoshi; Ota, Masayuki
Fusion Engineering and Design, 98-99, p.2178 - 2181, 2015/10
Times Cited Count:8 Percentile:53.98(Nuclear Science & Technology)We have analyzed the iron and concrete shielding experiments with the 40 and 65 MeV neutron sources at TIARA in Japan Atomic Energy Agency with the latest high-energy nuclear data libraries, JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0. The Monte Carlo code MCNP-5 and ACE files of JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0, which were supplied from JAEA, BNL and IAEA, respectively, were used for this analysis. The followings are found out from the results. (1) The calculations with JENDL/HE-2007 agree with all the measured ones well; (2) Those with ENDF/B-VII.1 tend to overestimate the measured ones with the thickness of the assemblies largely; (3) Those with FENDL-3.0 agree with the measured ones well for the iron experiment, while they overestimate the measured ones well for the concrete experiment largely. Some data in ENDF/B-VII.1 and FENDL-3.0 should be revised.
Ota, Masayuki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara
Fusion Engineering and Design, 98-99, p.1847 - 1850, 2015/10
Times Cited Count:2 Percentile:16.69(Nuclear Science & Technology)International Reactor Dosimetry and Fusion File release 1.0 (IRDFF 1.0), has been released from the International Atomic Energy Agency (IAEA) recently. In order to validate and test IRDFF 1.0, IAEA has initiated a new Co-ordinated Research Project (CRP). Under this CRP, we have performed an integral experiment on a graphite pseudo-cylindrical slab assembly with DT neutron source at JAEA/FNS. The graphite assembly of 31.4 cm in equivalent radius and 61 cm in thickness is placed at a distance of about 20 cm from the DT neutron source. A lot of foils for the dosimetry reactions in IRDFF1.0 are inserted into the small spaces between the graphite blocks along the center axis of the assembly. After DT neutron irradiation, reaction rates for the dosimetry reactions are measured by the foil activation technique. This experiment is analyzed by using Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0. The experimental assembly and DT neutron source are modeled precisely in the MCNP calculation. The reaction rates calculated with IRDFF 1.0 as the response functions for the dosimetry reactions are compared with the experimental values. Also the calculations with JENDL Dosimetry File 99 (JENDL/D-99) are performed for comparison. The results calculated with IRDFF 1.0 show good agreement with the experimental results.
Konno, Chikara; Ota, Masayuki; Kwon, Saerom; Ochiai, Kentaro; Sato, Satoshi
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 4 Pages, 2015/05
For fusion reactor nuclear analyses we produce new nuclear group constant sets, we have produced new nuclear group constant sets similar with FUSION-J3 from JENDL-4.0 and FENDL-3.0. The materials in these sets are H, H, He, He, Li, Li, Be, B, B, C, N, O, F, Na, Mg, Al, Si, P, S, K, Ca, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Nb, Mo, Cd, W, Pb, Bi, Cl, Ta, Sn and Ga. The nuclear group constant sets of JENDL-4.0 and FENDL-3.0, FUSION-J40 and FUISON-F30 (neutron: 175 groups, : 42 groups, P5 approximation), were produced from MATXS files of JENDL-4.0 and FENDL-3.0, which were newly processed with the NJOY99 code, with the TRANSX code. KERMA, DPA and gas production libraries were also prepared from the MATXS files with TRANSX. Several test calculations are carried out in order to validate these nuclear group constant sets. They suggest that these group constant sets have no problem.
Hashimoto, Kazuyuki; Nagai, Yasuki; Kawabata, Masako; Sato, Nozomi*; Hatsukawa, Yuichi; Saeki, Hideya; Motoishi, Shoji*; Ota, Masayuki; Konno, Chikara; Ochiai, Kentaro; et al.
Journal of the Physical Society of Japan, 84(4), p.043202_1 - 043202_4, 2015/04
Times Cited Count:8 Percentile:51.89(Physics, Multidisciplinary)Hoshino, Tsuyoshi; Ochiai, Kentaro; Edao, Yuki; Kawamura, Yoshinori
Fusion Science and Technology, 67(2), p.386 - 389, 2015/03
Times Cited Count:13 Percentile:71.17(Nuclear Science & Technology)Demonstration power reactors (DEMOs) require advanced tritium breeders that have high stability at high temperatures. Therefore, the pebble fabrication of LiTiO with excess Li (LiTiO) as an advanced tritium breeder was carried out. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed. DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. The LiTiO pebbles exhibited good tritium release properties similar to the LiTiO pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water.
Konno, Chikara; Ota, Masayuki; Asahara, Hiroo; Ochiai, Kentaro; Sato, Satoshi
JAEA-Conf 2014-002, p.160 - 166, 2015/02
IAEA released a new version of Evaluated Nuclear Data Library (FENDL), FENDL-3.0 in 2012 in order to extend the neutron energy range of neutron-induced reactions from 20 MeV to more than 60 MeV and to include general purpose and activation data libraries for proton- and deuteron-induced reactions up to more than 60 MeV. We already reported the benchmark tests of the general purpose data library for neutron-induced reactions below 20 MeV in FENDL-3.0. Now we present the benchmark tests of the general purpose data library for neutron-induced reactions in FENDL-3.0 by using iron and concrete shielding experiments with the 40 and 65 MeV neutron sources at TIARA in JAEA. As a result, it is found out that the calculations with FENDL-3.0 agree with the measured ones for the iron experiment well, while they overestimate the measured ones for the concrete experiment more for the thicker assemblies.
Shinohara, Koji; Ishii, Keiichi*; Ochiai, Kentaro; Baba, Mamoru*; Sukegawa, Atsuhiko; Sasao, Mamiko*; Kitajima, Sumio*
Review of Scientific Instruments, 85(11), p.11E823_1 - 11E823_4, 2014/11
Times Cited Count:0 Percentile:0.00(Instruments & Instrumentation)Ochiai, Kentaro; Kawamura, Yoshinori; Hoshino, Tsuyoshi; Edao, Yuki; Takakura, Kosuke; Ota, Masayuki; Sato, Satoshi; Konno, Chikara
Fusion Engineering and Design, 89(7-8), p.1464 - 1468, 2014/10
Times Cited Count:8 Percentile:51.95(Nuclear Science & Technology)We have performed the tritium recovery experiment on fusion reactor blanket with DT neutrons at the Fusion Neutronics Source facility in Japan Atomic Energy Agency. The candidate breeding material, LiTiO pebble, was put into the container which was set up it into an assembly simulating water cooled ceramic breeding (WCCB) blanket. Helium sweep gas including H (1%) and/or HO (1%) was flowed and extracted tritium was collected to water bubblers during DT neutron irradiation. The LiTiO pebble was also heated up to a constant temperature at 573, 873 and 1073 K, respectively. We arranged the tritium recovery system to measure tritiated water moisture and tritium gas, separately, and to investigate the amount of recovered tritium and the chemical form. From our experiments, it was showed that the amount of recovered tritium was corresponded to the calculation value and the ratio of chemical form depended to the temperature and kinds of sweep gas.