Ota, Masayuki; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Konno, Chikara
Fusion Engineering and Design, 89(9-10), p.2164 - 2168, 2014/10
Titanium is contained in lithium titanate, which is a tritium breeding material candidate. In the nuclear design, accurate nuclear data are needed. However, few benchmark experiments had been performed for titanium. Thus we performed a benchmark experiment with a titanium assembly and a DT neutron source at JAEA/FNS. The assembly was a titanium slab of 45 cm45 cm40 cm covered with 5 or 10 cm thick LiO blocks. Dosimetry reaction rates were measured by the foil activation method of niobium, aluminum, indium, gold and tungsten inside the assembly. And fission rates of U and U were measured by using micro fission chambers. This experiment was analyzed by using Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.0, ENDF/B-VII.1, FENDL-2.1, JEFF-3.1.2, JENDL-3.3, JENDL-4.0, and JENDL-4.0u. The calculation results were compared with the measured one in order to validate nuclear data libraries of titanium.
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
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.
Konno, Chikara; Ota, Masayuki; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi
Fusion Engineering and Design, 89(9-10), p.1889 - 1893, 2014/10
A nuclear data library is one of the most important data which control the calculation accuracy in nuclear analyses for fusion reactor designs. IAEA compiles the best data from the evaluated nuclear data libraries in the world each nuclei for fusion reactor applications. This library is Fusion Evaluated Nuclear Data Library (FENDL). In 2008 IAEA started a new coordinate research project to update the current version FENDL-2.1, to extend the neutron energy range from 20 MeV more than 60 MeV for 180 nuclei. This new version FENDL-3.0 was released in 2012. We have benchmarked FENDL-3.0 with integral experiments at FNS in JAEA and TOF experiments at OKTAVIAN in Osaka University in Japan. The Monte Carlo code MCNP-5 and the ACE file of FENDL-3.0 supplied from IAEA were used for the benchmark test. We also carried out calculations with the current version FENDL-2.1 and the latest JENDL-4.0 for comparison. As a result, it is demonstrated that the FENDL-3 is as accurate as FENDL-2.1 or more.
Konno, Chikara; Ochiai, Kentaro; Takakura, Kosuke; Sato, Satoshi
Nuclear Data Sheets, 118, p.450 - 452, 2014/04
KERMA (Kinematic Energy Release in Material) factors are used as response data in order to obtain nuclear heating in nuclear analyses. These data are not directly included in nuclear data libraries and they in ACE (A Compact ENDF) files for the Monet Carlo radiation transport code MCNP are deduced from cross section data for all the reactions in nuclear data libraries with the NJOY code. Many peoples use these data unquestioningly, but little is known concerning fact that most of these data are not always correct. Thus we have examined KERMA factors in the latest official ACE files; those of JENDL-4.0, ENDF/B-VII.1, JEFF-3.1.1, etc. It is found out that a lot of KERMA data in the official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.1.1 are not always correct.
Sato, Satoshi; Maegawa, Toshio*; Yoshimatsu, Kenji*; Sato, Koichi*; Nonaka, Akira*; Takakura, Kosuke; Ochiai, Kentaro; Konno, Chikara
Progress in Nuclear Science and Technology (Internet), 4, p.623 - 626, 2014/04
In the previous study, we developed a multi-layered concrete structure to reduce induced activity in concrete applied for neutron generation facilities such as a fusion reactor. This structure is composed of low activation concrete as the first layer, boron doped low activation concrete as the second layer and ordinary concrete as the third layer from the side of the neutron source. In this study, as an alternative of the boron doped low activation concrete we have developed the boron doped resin sheet with boron carbonate and resin to reduce the construction cost. The weight ratio of the boron carbonate to the resin is 0.75. The developed boron sheet has good flexibility and sufficient strength for repeated bending. DT neutron irradiation experiments for four multi-layered concrete structures with the boron sheet have been performed at the FNS (Fusion Neutronics Source) facility in JAEA in order to study shielding performance of the structures with the boron sheet. Structure-1 of about 30 cm in width, 30 cm in height and 50 cm in thickness is composed of low activation concrete of 20 cm in thickness as the first layer and ordinary concrete of 30 cm in thickness as the second layer. The boron sheet is inserted between the first and second layers. In Structure-2 one more boron sheet is inserted at the 10 cm depth from the surface of Structure-1. Structure-3 added one more boron sheet at 30 cm depth from the surface of Strucure-2. For comparison, Structure-4 has no boron sheet. The reaction rates were measured every 5 cm in depth with activation foils of gold and niobium. By inserting the boron sheet, the reaction rate of the gold generated by low energy neutrons decreases by a factor of about four. It is demonstrated that the multi-layered concrete structure with the boron sheet effectively reduces low energy neutrons.
Konno, Chikara; Kato, Yoshinari*; Takakura, Kosuke; Ota, Masayuki; Ochiai, Kentaro; Sato, Satoshi
Progress in Nuclear Science and Technology (Internet), 4, p.606 - 609, 2014/04
At International Conference on Nuclear Data for Science and Technology in 2007 we pointed out that most of unresolved resonance data in JENDL-3.3 have a problem related to self-shielding correction. Here with a simple calculation model we investigated if the newest JENDL, JENDL-4.0, was improved for the problem or not. As a result, it seems that unresolved resonance data in JENDL-4.0 have no problem, but we are afraid that the self-shielding effect for the unresolved resonance data in JENDL-4.0 is too large. New integral experiments for unresolved resonance data are strongly recommended in order to verify unresolved resonance data.
Kato, Yoshinari*; Takakura, Kosuke; Ota, Masayuki; Ochiai, Kentaro; Sato, Satoshi; Konno, Chikara
Progress in Nuclear Science and Technology (Internet), 4, p.596 - 600, 2014/04
We have performed benchmark tests for JENDL-4.0 released last year in shielding and fusion neutronics fields. Now we analyze OKTAVIAN TOF experiments (CF, Si, Ti, Cr, Mn, Co, Cu, As, Se, Zr, Nb, Mo, W) with JENDL-4.0 in order to validate JENDL-4.0. For comparison we also do with the older version JENDL-3.3 and other recent nuclear data libraries (ENDF/B-VII.0, JEFF-3.11). The Monte Carlo code MCNP-4C was used for these analyses. We adopted the official ACE files for JENDL-4.0, JENDL-3.3, JEFF-3.1 and ENDF/B-VII.0. As a result, the following results are obtained through comparison between calculation and measured results. (1) Si, As, Se, Mo, W : Calculation results with JENDL-4.0 agree with the measured ones better than those with JENDL-3.3. (2) CF, Co, Cu, Ti, Zr : Calculation results with JENDL-4.0 are almost the same as those with JENDL-3.3. (3) Cr, Mn, Nb : Calculation results with JENDL-4.0 are partially better and partially worse than those with JENDL-3.3.
Ota, Masayuki; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Konno, Chikara
Fusion Engineering and Design, 88(9-10), p.2160 - 2163, 2013/10
Three-dimensional Monte Carlo transport code is widely used in analyses of radiation protection and shielding, nuclear criticality safety, fission and fusion reactor design and nuclear instrumentation. The MCNP code is worldwide used as the standard code. The TRIPOLI code is adopted as one of ITER neutronics codes, but it is not always used worldwide, particularly in Japan. In this study we examined the basic performance of TRIPOLI through simple model calculations and analysis of iron neutronics experiments with DT neutrons at the Fusion Neutronics Source facility in Japan Atomic Energy Agency. Almost results by TRIPOLI are consistent with ones by MCNP. However, there is a difference between TRIPOLI and MCNP results in some calculations caused mainly by treatment with inelastic scattering data.
Kin, Tadahiro*; Nagai, Yasuki; Iwamoto, Nobuyuki; Minato, Futoshi; Iwamoto, Osamu; Hatsukawa, Yuichi; Segawa, Mariko; Harada, Hideo; Konno, Chikara; Ochiai, Kentaro; et al.
Journal of the Physical Society of Japan, 82(3), p.034201_1 - 034201_8, 2013/03
We have measured the activation cross sections producing Cu and Cu, promising medical radioisotopes for molecular imaging and radioimmunotherapy, by bombarding a natural zinc sample with 14 MeV neutrons. We estimated the production yields of Cu and Cu by fast neutrons from C(d,n) with 40 MeV 5 mA deuterons. The calculated Cu yield is 1.8 TBq (175 g Zn) for 12 h of irradiation; the yields of Cu by Zn(n,p)Cu and Zn(n,x)Cu were 249 GBq (184 g Zn) and 287 GBq (186 g Zn) at the end of 2 days of irradiation, respectively. From the results, we proposed a new route to produce Cu with very little radionuclide impurity via the Zn(n,x)Cu reaction, and showed the Zn(n,p)Cu reaction to be a promising route to produce Cu.
Edao, Yuki; Kawamura, Yoshinori; Ochiai, Kentaro; Hoshino, Tsuyoshi; Takakura, Kosuke; Ota, Masayuki; Iwai, Yasunori; Yamanishi, Toshihiko; Konno, Chikara
JAEA-Research 2012-040, 15 Pages, 2013/02
Tritium generation and recovery studies on LiTiO as a solid breeding material under neutron irradiation carried out in the Fusion Neutron Source (FNS) facility. A capsule with LiTiO packed bed was put in a system which simulated an actual blanket system which built in beryllium blocks and lithium titanate ones. Estimated values of the amount of tritium generation by a numerical calculation agreed closely with experimental values. The capsule was heated up to 300C, and helium, helium with water vapor, hydrogen or hydrogen/water vapor were selected as purge gas. In the case of purge by helium added water vapor, the ratio of HTO to total tritium release was 98%. In helium with hydrogen/water vapor purge, the ratio of HTO to total tritium release was 80%, which was confirmed that HTO released by isotope exchange reaction between water vapor and tritium. In helium with hydrogen purge, the ratio of HT to total tritium release was 6070%, which was shown that HT released by isotope exchange reaction between hydrogen gas and tritium. HTO released by water generation reaction between hydrogen in purge gas and oxygen in LiTiO although water vapor was not added in purge gas. The ratio of HTO release seemed to be small under the deoxidized condition of the LiTiO surface. Tritium release behavior in the LiTiO depended on the composition of purge gas, and its chemical form was affected by the surface conditions of LiTiO.
Onishi, Seiki*; Kondo, Keitaro*; Azuma, Tetsushi*; Sato, Satoshi; Ochiai, Kentaro; Takakura, Kosuke; Murata, Isao*; Konno, Chikara
Fusion Engineering and Design, 87(5-6), p.695 - 699, 2012/08
A new integral experiment with a deuteron-triton fusion (DT) neutron beam started in order to validate scattering cross section data. First the DT neutron beam was constructed with a collimator. The characteristics of the DT neutron beam were examined experimentally. Second a new integral experiment for type 316 stainless steel (SS316) was carried out with this DT neutron beam. Reaction rates of the Nb(n,2n)Nb reaction on the center of the beam axis and at 15 cm and 30 cm apart from the axis in the assembly were measured with the activation foil method and were calculated with the Monte Carlo transport calculation code MCNP and nuclear data libraries, JENDL-4.0, JENDL-3.3 and ENDF/B-VI.8. The ratios of calculation to experiment became smaller than 1 with the distance from the beam axis for all the nuclear libraries. It was pointed out that the diagonally forward cross section data had some problems.
Kondo, Keitaro; Yagi, Takahiro*; Ochiai, Kentaro; Sato, Satoshi; Takakura, Kosuke; Onishi, Seiki; Konno, Chikara
Fusion Engineering and Design, 86(9-11), p.2184 - 2187, 2011/10
In the neutronics experiment for the ITER test blanket module with a Li-enriched LiTiO layer and a beryllium layer conducted at the FNS facility of Japan Atomic Energy Agency, the calculated tritium production rate (TPR) was by approximately 10% larger than the measured one only when a neutron source reflector composed of SS316 was attached. On the other hand, the influence of the reflector on the TPR prediction accuracy was not seen in the recent blanket experiment with a natural LiTiO layer, beryllium layers and the reflector. We investigated the former experiment in detail, and found an unphysical tendency in the measured TPR distribution. In order to clarify whether the deterioration of the TPR prediction accuracy originates from the reflector or not, we have conducted the same experiment as the previous experiment again. In the present experiment, the measured TPR distribution inside the Li-enriched LiTiO layer well agreed with the calculated one within an estimated experimental error of 6%. We conclude that the overestimation of TPR observed in the previous experiment would be due to some experimental errors and that the TPR prediction accuracy is good even in the case with the reflector.
Konno, Chikara; Wada, Masayuki*; Kondo, Keitaro; Onishi, Seiki; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi
Fusion Engineering and Design, 86(9-11), p.2682 - 2685, 2011/10
JENDL-4, the major revised version of Japanese Evaluated Nuclear Data Library (JENDL), was released in spring, 2010. We analyzed the fusion neutronics benchmark experiments on iron at JAEA/FNS with JENDL-4.0 and MCNP4C as the detail benchmark test of JENDL-4.0 iron data. As a result, it is found out that the problems of iron data in JENDL-3.3 are adequately revised in JENDL-4.0 iron data; e.g. the first inelastic scattering cross section data of Fe and angular distribution of elastic scattering of Fe. The iron data in JENDL-4.0 are comparable to and are partly better than those in ENDF/B-VII.0 and JEFF-3.1.
Sato, Satoshi; Maegawa, Toshio*; Yoshimatsu, Kenji*; Sato, Koichi*; Nonaka, Akira*; Takakura, Kosuke; Ochiai, Kentaro; Konno, Chikara
Journal of Nuclear Materials, 417(1-3), p.1131 - 1134, 2011/10
The multi-layered concrete structure has been developed to reduce induced activity in the concrete applied for neutron generation facilities such as a fusion reactor. The multi-layered concrete structure is composed of the low activation concrete as the first layer, the boron-doped low activation concrete as the second layer and the ordinary concrete as the third layer from the side of the neutron source. By applying the multi-layered concrete structure, the volume of the boron can drastically decrease compared with the monolithic boron-doped concrete. A 14 MeV neutron irradiation experiment with the multi-layered concrete structure mockups was performed at FNS and several reaction rates and induced activities in the mockups were measured. This experiment demonstrated that the multi-layered concrete effectively reduced low energy neutrons and induced activities.
Konno, Chikara; Takakura, Kosuke; Kondo, Keitaro; Onishi, Seiki*; Ochiai, Kentaro; Sato, Satoshi
Progress in Nuclear Science and Technology (Internet), 2, p.341 - 345, 2011/10
We have already pointed out that the background cross sections and weighting flux are not adequate in multigroup libraries VITAMIN-B6 and JSSTDL-300. This time we examined if the latest multigroup libraries, MATXS-J33, MATJEFF3.1.BOLIB, VITJEFF3.1. BOLIB, VITENEA-J, HILO2k and AMPX file of ENDF/B-VII.0 in SCALE6 and MTXS file in ADS-2.0 have the same problems. The followings are found out from our simple calculations. (1) MATXS-J33 has no problem. (2) VITJEFF3.1.BOLIB, VITENEA-J, HILO2k and AMPX file of ENDF/B-VII.0 in SCALE6 are produced by using the inadequate weighting flux. (3) VITJEFF3.1.BOLIB, MATJEFF3.1.BOLIB and MTXS file in ADS-2.0 have inadequate background cross sections. Note that the self-shielding correction in calculations with VITJEFF3.1.BOLIB, VITENEA-J, HILO2k and AMPX file of ENDF/B-VII.0 in SCALE6, MATJEFF3.1.BOLIB and MTXS file in ADS-2.0 is not always adequate.
Konno, Chikara; Takakura, Kosuke; Wada, Masayuki*; Kondo, Keitaro; Onishi, Seiki*; Ochiai, Kentaro; Sato, Satoshi
Progress in Nuclear Science and Technology (Internet), 2, p.346 - 357, 2011/10
The major revised version of Japanese Evaluated Nuclear Data Library (JENDL), JENDL-4, was released in 2010 spring. As the benchmark test of JENDL-4.0 in the shielding and fusion neutronics fields, we analyzed many integral benchmark experiments (in-situ and Time-of-Flight (TOF) experiments) with DT neutrons at JAEA/FNS with the MCNP code and JENDL-4.0. The experiments with assemblies including beryllium, carbon, silicon, vanadium, copper, tungsten and lead, nuclear data of which were revised in JENDL-4.0, were selected for this benchmark test. As a result, it is found that JENDL-4 improved some problems pointed out in JENDL-3.3 and that it is comparable to ENDF/B-VII.0 and JEFF-3.1.
Hatsukawa, Yuichi; Nagai, Yasuki; Kin, Tadahiro; Segawa, Mariko; Harada, Hideo; Iwamoto, Osamu; Iwamoto, Nobuyuki; Ochiai, Kentaro; Takakura, Kosuke; Konno, Chikara; et al.
Proceedings in Radiochemistry, 1(1), p.327 - 329, 2011/09
Authors proposed a new route to produce Mo by the Mo(,2)Mo reaction, which has some characteristic features. Such as the reaction cross section is large, about 1.5 barn at 12 17 MeV, which is 10 times larger than the thermal-neutron capture cross section of Mo. Second, the cross sections of the (,), (,), and (,) reactions are less than a few mb at = 14 MeV. Third, a large amount of Mo target materials can be used, compared to that for proton beam irradiation on Mo. Fourth, intense neutrons with energy of 12-17 MeV are already available. In the present work we have measured all -rays emitted from activities produced by bombarding a natural Mo target with neutrons from the D(H,)He reaction at Fusion Neutronics Source Facility (FNS) at Japan Atomic Energy Agency (JAEA) to study characteristic features mentioned above more in detail. The neutron flux was about 10n/cms. The experimental results at FNS will be discussed in the conference.
Tanaka, Teruya*; Sato, Satoshi; Kondo, Keitaro; Ochiai, Kentaro; Murata, Isao*; Takakura, Kosuke; Sato, Fuminobu*; Kada, Wataru*; Iida, Toshiyuki*; Konno, Chikara; et al.
Fusion Science and Technology, 60(2), p.681 - 686, 2011/08
Irradiation experiments of 14 MeV neutrons have been performed on a Li block assembly of 46 51 51 cm with a 5 cm thick V-alloy layer inside to examine the accuracy of neutronics calculations for the Li/V-alloy blanket design. Foils of Nb, Ni, In and Au for reaction rate measurements of Nb(n,2n)Nb, Ni(n,p)Co, In(n,n')In, Au(n,)Au reactions and Li enriched (Li: 95.5%) and Li enriched (Li: 99.9%) LiCO pellets for tritium production rate measurements were installed in the assembly. Results of the measurements were compared with those of calculations with MCNP5, JENDL-3.3 and JENDL/D-99. The comparisons for the reaction rates in the Nb, Ni and In foils indicate that measurements and calculations of the fast neutron transport are consistent almost within 10%. In the comparison for the reaction rates in the Au foils, the underestimation of 15% was found at a surface of the V-alloy layer. There is a possibility that this is due to the elastic scattering cross section of V around 4 keV as previously reported. The comparisons for tritium production rates in the Li enriched and Li enriched LiCO pellets indicate that calculated rates were larger than results of the measurements by 2-8% and 1-4%, respectively.
Onishi, Seiki; Kondo, Keitaro; Sato, Satoshi; Ochiai, Kentaro; Takakura, Kosuke; Konno, Chikara; Murata, Isao*
Journal of the Korean Physical Society, 59(2), p.1949 - 1952, 2011/08
So far we carried out many integral benchmark experiments (in-situ experiments and Time-Of-Flight experiments) for nuclear data with DT neutrons at the Fusion Neutronics Source facility in Japan Atomic Energy Agency. In addition to those, we have a plan to perform new integral benchmark experiments for nuclear data with a DT neutron beam, which can investigate nuclear data for almost the whole angle and the whole energy. Because the large-size tritium target in FNS is difficult to procure, we have started to build a new DT neutron beam with the small tritium target, which is easy to obtain, at the first target room of FNS. We already designed a collimator system for the DT neutron beam based on calculations. In this work, under that design, the collimator was constructed. Then the characteristics of the neutron field were measured in order to confirm the DT neutron beam performance. It was demonstrated that the DT neutron beam was realized as calculated.
Ochiai, Kentaro; Kondo, Keitaro; Onishi, Seiki; Takakura, Kosuke; Sato, Satoshi; Abe, Yuichi; Konno, Chikara; Suzuki, Chihiro*; Yagi, Takahiro*
Journal of the Korean Physical Society, 59(2), p.1953 - 1956, 2011/08
Lead is an important candidate material as multiplier of nuclear fusion reactor. Few DT neutron integral benchmark experiments were performed for lead so far. Therefore, we have carried out an integral benchmark experiment on lead at the DT neutron source facility of JAEA, FNS. A cubic lead assembly on a side of 45.3 cm was set up and was irradiated with the DT neutron source. Reaction rates of the Al(n,)Na, Nb(n,2n)Nb, Zr(n,2n)Zr and In(n,n')In reactions were measured as fast neutron spectrum indices in the assembly. A small NE213 spectrometer was also used for measurement of neutron spectra in the assembly. A Monte Carlo calculation code, MCNP5, was adopted to calculate the above neutron spectra and activation reaction rates. Nuclear data libraries, JENDL-3.3, ENDF/B-VII.0, JEFF-3.1 and FENDL-2.1, were used in the calculation. The calculation results of the three libraries except for JENDL-3.3 agreed with the measuring ones. In case of JENDL-3.3, some remarkable disagreements were found. From our investigations, it was pointed out that the inappropriate evaluation of the (n,2n) and inelastic cross sections of lead in JENDL-3.3 caused such disagreement.