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Journal Articles

TENDL-2017 benchmark test with iron shielding experiment at QST/TIARA

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 $$^{54}$$Fe, $$^{56}$$Fe and $$^{58}$$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.

Journal Articles

Effect of IAEA patch for TRANSX2.15

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.

Journal Articles

FENDL-3.1b test

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 $$^{15}$$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 $$^{116}$$Sn and $$^{117}$$Sn NJOY processing at IAEA and specified that one of NJOY patches caused this trouble.

Journal Articles

ENDF/B-VIII$$beta$$2 benchmark test with shielding experiments at QST/TIARA

Kwon, Saerom*; Konno, Chikara; Ota, Masayuki*; Sato, Satoshi*; Ochiai, Kentaro*

JAEA-Conf 2017-001, p.123 - 128, 2018/01

The $$beta$$-version of ENDF/B-VIII, ENDF/B-VIII$$beta$$2, was released in August, 2016. Thus we studied whether the overestimation problems due to the $$^{16}$$O and $$^{56}$$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-VIII$$beta$$2 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 5$$^{56}$$Fe data was improved, (2) the overestimation for around 65 MeV due to the $$^{56}$$Fe data was also slightly improved, though it was worse than that with FENDL-3.1b, (3) the drastic overestimation due to the $$^{16}$$O data was not improved. The final version of ENDF/B-VIII should also be modified based on these results.

Journal Articles

Benchmark experiment on copper with graphite by using DT neutrons at JAEA/FNS

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:64.68(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.

Journal Articles

Lead benchmark experiment with DT neutrons at JAEA/FNS

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:64.68(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 $$gamma$$ 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$$_{2}$$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.

Journal Articles

Important comments on KERMA factors and DPA cross-section data in ACE files of JENDL-4.0, JEFF-3.2 and ENDF/B-VII.1

Konno, Chikara; Tada, Kenichi; Kwon, Saerom*; Ota, Masayuki*; Sato, Satoshi*

EPJ Web of Conferences (Internet), 146, p.02040_1 - 02040_4, 2017/09

 Percentile:100

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 $$gamma$$ data, wrong secondary $$gamma$$ 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.

Journal Articles

JENDL-4.0/HE benchmark test with concrete and iron shielding experiments at JAEA/TIARA

Konno, Chikara; Matsuda, Norihiro; Kwon, Saerom*; Ota, Masayuki*; Sato, Satoshi*

EPJ Web of Conferences (Internet), 153, p.01024_1 - 01024_6, 2017/09

 Times Cited Count:1 Percentile:14.04

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 $$^{56}$$Fe caused the underestimation in the calculation with JENDL-4.0/HE for the iron experiment.

Journal Articles

Benchmark experiment on molybdenum with graphite by using DT neutrons at JAEA/FNS

Ota, Masayuki*; Kwon, Saerom*; Sato, Satoshi*; Konno, Chikara; Ochiai, Kentaro*

Fusion Engineering and Design, 114, p.127 - 130, 2017/01

 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,$$gamma$$) cross section of $$^{95}$$Mo is underestimated in the tail region below the large resonance at 45 eV in the recent nuclear data libraries.

Journal Articles

New remarks on KERMA factors and DPA cross section data in ACE files

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.

Journal Articles

New integral experiments for a variety of fusion reactor materials with DT neutron source at JAEA/FNS

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 Li$$_{2}$$O 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.

Journal Articles

Some comments on KERMA factors and DPA cross-section data in ACE and MATXS files of JENDL-4.0

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) $$gamma$$ 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.

Journal Articles

A Simple method for modification of capture reaction and elastic scattering nuclear data in analyses of nuclear data benchmark experiments

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.

Journal Articles

Problems on FENDL-3.0

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 $$^{16}$$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.

Journal Articles

JENDL/HE-2007 benchmark test with iron shielding experiment at JAEA/TIARA

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 $$^{56}$$Fe in JENDL/HE-2007 caused the underestimation. The non-elastic scattering data of $$^{56}$$Fe in JENDL/HE-2007 should be revised.

Journal Articles

New nuclear data group constant sets for fusion reactor nuclear analyses based on FENDL-2.1

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, $$gamma$$: 42 groups, P5 approximation) and FUSION-F21.42 (neutron: 42 groups, $$gamma$$: 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.

Journal Articles

New nuclear data group constant sets for fusion reactor nuclear analyses based on JENDL-4.0 and FENDL-3.0

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 $$^{1}$$H, $$^{2}$$H, $$^{3}$$He, $$^{4}$$He, $$^{6}$$Li, $$^{7}$$Li, $$^{9}$$Be, $$^{10}$$B, $$^{11}$$B, $$^{12}$$C, $$^{14}$$N, $$^{16}$$O, $$^{19}$$F, $$^{23}$$Na, Mg, $$^{27}$$Al, Si, $$^{31}$$P, S, K, Ca, Ti, $$^{51}$$V, Cr, $$^{55}$$Mn, Fe, Co, Ni, Cu, Zr, $$^{93}$$Nb, Mo, Cd, W, Pb, $$^{209}$$Bi, Cl, $$^{181}$$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, $$gamma$$: 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.

Oral presentation

Design modification related to pressure resistance for water-cooled ceramic breeder blanket

Tanigawa, Hisashi; Sato, Satoshi; Kwon, Saerom; Enoeda, Mikio

no journal, , 

For development of water-cooled solid breeder blanket, design of container structure of the blanket is modified to improve pressure resistance property. The higher temperature and pressure of the coolant water are preferable for efficiency in electric power generation. On the other hand, the smaller amount of the structural material is preferable in view of tritium breeding capability. The coolant pipe break leads to inner pressure loading on the container structure. To improve the pressure resistance property of the container, geometrical conditions of the container has been studied.

Oral presentation

Benchmark experiment on molybdenum with DT neutrons at JAEA/FNS

Ota, Masayuki; Kwon, Saerom; Ochiai, Kentaro; Sato, Satoshi; Konno, Chikara

no journal, , 

We perform a benchmark experiment with a Mo assembly and the DT neutron source at JAEA/FNS to validate recent nuclear data of Mo. A rectangular Mo assembly, the size of which is 253 mm $$times$$ 253 mm $$times$$ 354 mm, is covered with 51, 202 and 253 mm thick Li2O blocks around the front, side and back surfaces in order to eliminate background neutrons in the measuring points, respectively. The assembly is placed at a distance of 150 mm from the DT neutron source. Several dosimetry reaction rates and fission rates measured in the assembly are compared to those calculated with the Monte Carlo neutron transport code MCNP5-1.40 and the recent nuclear data libraries of ENDF/B-VII.1, JEFF-3.2 and JENDL-4.0 (FENDL-3.0). The ratios of the calculated reaction rates to the experimental ones generally decrease with the increasing distance from the front surface of the assembly. Reasons of the discrepancies are discussed in the presentation.

Oral presentation

A New benchmark experiment on copper with DT neutron source at JAEA/FNS

Kwon, Saerom; Ota, Masayuki; Ochiai, Kentaro; Sato, Satoshi; Konno, Chikara

no journal, , 

A benchmark experiment on copper with DT neutron source was performed 20 years ago at JAEA/FNS. However, the calculated results tended to underestimate the measured data related to lower energy neutrons below a few keV in the experiment, which suggested that the measured data might be affected by neutrons scattered in the concrete wall of the experiment room or other surroundings. Therefore, we have carried out an additional integral experiment on copper, where a copper assembly was covered with Li$$_{2}$$O blocks to reduce neutrons scattered in the concrete wall. We used the Monte Carlo neutron transport code, MCNP5-1.40 and the recent nuclear data libraries, ENDF/B-VII.1, JEFF-3.2, JENDL-4.0 and FENDL-3.0 (ENDF/B-VII.0) for the experiment analysis. JENDL/D-99 was used as dosimetry cross section data. The calculated reaction rates of the $$^{197}$$Au(n,$$gamma$$)$$^{198}$$Au reaction with all the nuclear data libraries still underestimate the measured data, although the underestimation was improved compared to the previous result with JENDL-4.0 in particular. We found out that the nuclear data of copper caused this underestimation problem.

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