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Sato, Satoshi; Iida, Hiromasa; Ochiai, Kentaro; Konno, Chikara; Nishitani, Takeo; Morota, Hidetsugu*; Nashif, H.*; Yamada, Masao*; Masuda, Fukuzo*; Tamamizu, Shigeyuki*; et al.
Nuclear Technology, 168(3), p.843 - 847, 2009/12
Times Cited Count：7 Percentile：45.16(Nuclear Science & Technology)It takes huge or unrealistic amounts of time to prepare accurate calculation inputs in shielding design for very large and complicated structure such as fusion reactors. For that reason, we have developed an automatic conversion system from three dimensional CAD drawing data into input data of the calculation geometry for a three dimensional Monte Carlo radiation transport calculation code MCNP, and applied it to an ITER benchmark model. This system consists of a void creation program (CrtVoid) for CAD drawing data and a conversion program (GEOMIT) from CAD drawing data to MCNP input data. CrtVoid creates void region data by subtracting solid region data from the whole region by Boolean operation. The void region data is very large and complicated geometry. The program divides the overall region to many small cubes, and the void region data can be created in each cube. GEOMIT generates surface data for MCNP data based on the CAD data with voids. These surface data are connected, and cell data for MCNP input data are generated. In generating cell data, additional surfaces are automatically created in the program, and undefined space and duplicate cells are removed. We applied this system to the ITER benchmark model. We successfully created void region data, and MCNP input data. We calculated neutron flux and nuclear heating. The calculation results agreed well with those with MCNP inputs generated from the same CAD data with other methods.
Masukawa, Fumihiro; Nakane, Yoshihiro; Iwamoto, Yosuke; Nakashima, Hiroshi
Nuclear Technology, 168(3), p.680 - 684, 2009/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The radioactivity produced in the accelerator cooling water was estimated to draw up the maintenance scenario of the J-PARC accelerators. Activation cross section sets from Nitrogen and Oxygen were evaluated from the available experimental data and theoretically calculated data by INC/GEM and LAHET for high-energy proton and neutron. The PHITS and MCNPX codes were used to calculate the proton and neutron fluxes in the water-cooled accelerator components. The radioactivity from the corrosion product was also estimated by scaling of the measurements at KEK-PS and LAMPF. Tritium are estimated 5.7
10
[Bq] in LINAC and 2.510
[Bq] in 3-GeV Synchrotron by annual operation. Dose in machinery rooms by short-lived nuclides are reasonably low at LINAC, whereas 
10
times higher at 3-GeV Synchrotron. The estimation of tritium is acceptable level for the disposal to the environment, while short-lived nuclides at 3-GeV Synchrotron may make the facility maintenance difficult.
Sakamoto, Yukio; Iwamoto, Yosuke; Nakashima, Hiroshi
Nuclear Technology, 168(3), p.654 - 658, 2009/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)We review some accelerator benchmark experiments on source term and radiation shielding conducted by JAEA to validate the accuracy of the radiation behavior simulation code such as PHITS code. (1) Neutron spectra measurement from thin beryllium target bombarded with about 10 MeV protons for accelerator boron neutron capture therapy. (2) Forward direction neutron spectra measurements and shielding experiments from beryllium target bombarded with 140, 250 and 350 MeV protons. (3) Neutron spectra measurement from tungsten target bombarded with about 400 MeV protons for the establishment of irradiation field for semiconductor tests. (4) Shielding experiments at Pbar target station and NuMI target station of Fermi Lab.
Sakamoto, Yukio; Hirayama, Hideo*; Sato, Osamu*; Shimizu, Akinao*
Nuclear Technology, 168(3), p.585 - 590, 2009/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)A Bremsstrahlung production data is needed in the calculation of buildup factors included by the contribution of Bremsstrahlung as secondary photons by IE method. In this work, the emission of Bremsstrahlung is treated as possible as exactly by the introduction of EGS4 results. The Bremsstrahlung production data by pair-created electrons and Compton scattered electrons is evaluated for 26 elements and 4 compound and mixtures. The error estimation of Bremsstrahlung contribution to buildup factors by IE method coupled with this Bremsstrahlung data is coincident with fully transported results by EGS4 code within about 5%. By the introduction of this Bremsstrahlung production data into IE methods, we can calculate buildup factors included by the contribution of Bremsstrahlung with good accuracy up to deep penetration.
Konno, Chikara; Sato, Satoshi; Ochiai, Kentaro; Wada, Masayuki*; Onishi, Seiki; Takakura, Kosuke; Iida, Hiromasa
Nuclear Technology, 168(3), p.743 - 746, 2009/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The three-dimensional Sn code Attila of Transpire, Inc. can use CAD data as a geometrical input directly and deal with assemblies of complicated geometry without much effort. ITER organization has a plan to adopt this code as one of the standard codes for nuclear analyses. However validation of calculations with this code is not carried out in detail so far. Thus we validate this code through analyses of some bulk experiments and streaming experiments with DT neutrons at JAEA/FNS. Analyses with the Sn code system DOORS and Monte Carlo code MCNP4C were also carried out for comparison. Agreement between Attila and DOORS calculations is very good for the bulk experiments. For streaming experiments Attila requires special treatments (biased angular quadrature sets or last collided source calculation) as well as DOORS in order to obtain similar results as those with MCNP, though Attila consumes much more time and memory than DOORS.
Nakane, Yoshihiro; Abe, Teruo*; Nakashima, Hiroshi
Nuclear Technology, 168(2), p.519 - 523, 2009/11
Times Cited Count：1 Percentile：10.21(Nuclear Science & Technology)For the shielding design of 3-GeV synchrotron in J-PARC, radiation shielding calculations were performed around beam injection and beam extraction areas. It is difficult to apply the simplified methods for these areas that the geometries are complicated and the larger beam losses are assumed. Therefore, Monte Carlo code, PHITS, were applied to particle transport calculations for the shielding designs around those areas. The shielding designs were performed with the condition that the maximum dose rates at the boundary of radiation controlled areas are in less than both the design criterion defined in J-PARC and the limit defined in regulation of Japan. Calculation methods and conditions used for the shielding design and the calculation results are presented in this work.
Nakashima, Hiroshi; Sakamoto, Yukio; Iwamoto, Yosuke; Matsuda, Norihiro; Kasugai, Yoshimi; Nakane, Yoshihiro; Masukawa, Fumihiro; Mokhov, N.*; Leveling, A.*; Boehnlein, D.*; et al.
Nuclear Technology, 168(2), p.482 - 486, 2009/11
Times Cited Count：7 Percentile：45.16(Nuclear Science & Technology)Experimental studies of shielding and radiation effects have been started using 120-GeV proton synchrotron at Fermi National Accelerator Laboratory (FNAL) under collaboration between FNAL and Japan. The first campaign of the experiment was carried out at the Pbar target station and Numi experimental station at FNAL, using antiproton and neutrino production targets irradiated by 120-GeV protons. The generated secondary particles passing through steel, concrete and rock were measured by activation methods as well as by other detectors such as scintillator with a veto counter, phoswich detector and a Bonner ball counter on trial. Preliminary experimental results are presented.
Iwamoto, Yosuke; Satoh, Daiki; Hagiwara, Masayuki*; Iwase, Hiroshi*; Kirihara, Yoichi*; Yashima, Hiroshi*; Nakane, Yoshihiro; Nakashima, Hiroshi; Nakamura, Takashi*; Tamii, Atsushi*; et al.
Nuclear Technology, 168(2), p.340 - 344, 2009/11
Times Cited Count：5 Percentile：35.91(Nuclear Science & Technology)It is important to calculate neutron production from thick targets at forward angle near 0 degree and backward angle near 180 degree for the shielding design of proton accelerator facilities, and the accuracy should be determined by experimental data. There are, however, few experimental data near 0 degree, and no data near 180 degree in the energy region above 100 MeV. Neutron energy spectra at 0, 90 and 180 degrees produced from thick graphite and iron targets by 140 MeV protons were measured. It was found that the calculation with JENDL-HE was more suitable for the shielding design of proton accelerator facilities in the forward direction.
Sukegawa, Atsuhiko; Okuno, Koichi*; Sakurai, Shinji
Nuclear Technology, 168(2), p.553 - 558, 2009/11
Times Cited Count：4 Percentile：30.43(Nuclear Science & Technology)The resin as a neutron shielding material with flexibility in shape and lightness is expected to be set up near the fusion tokamak devices and as the collimator of the diagnostics. KRAFTON-HB4 is epoxy-based resin that contains boron. It has been developed for future FBR shielding materials. The heat resistant temperature is 
150
C. EPONITE is neutron shielding material based on Colemanite and epoxy-based resin that contains boric acid. It has been developed for a PET shielding materials. The heat resistant temperature is 200C. The boron-loaded resin with the heat resistance applicable to 300C has been newly developed. The developed resin is made of phenol-based resin that has improved heatproof. The weight percent of hydrogen and boron are the developed resin (H: 1.94, B: 6.1), KRAFTON-HB4 (H: 10.5, B: 2.0), and EPONITE (H: 3.8, B: 30.4). Neutron attenuation experiments of these resins were carried out using 
Cf neutron source. In the experiments, neutron dose attenuation were measured by a REM counter. The tenth layer of Polyethylene, the developed resin, KRAFTON-HB4 and EPONITE are 0.14 m, 0.14 m, 0.15 m and 0.09 m, respectively.
Sato, Tatsuhiko; Satoh, Daiki; Endo, Akira; Shigyo, Nobuhiro*; Yasuda, Hiroshi*; Takada, Masashi*; Yajima, Kazuaki*; Nakamura, Takashi
Nuclear Technology, 168(1), p.113 - 117, 2009/10
Times Cited Count：1 Percentile：10.21(Nuclear Science & Technology)The details of the features of DARWIN will be presented at the meeting, together with the recent improvements of the system such as the implementation of the function for estimating neutron spectra using the unfolding technique.
Saito, Kimiaki; Sato, Kaoru; Endo, Akira; Kinase, Sakae
Nuclear Technology, 168(1), p.213 - 219, 2009/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Kinase, Sakae; Matsuhashi, Shimpei; Saito, Kimiaki
Nuclear Technology, 168(1), p.154 - 157, 2009/10
Times Cited Count：4 Percentile：30.43(Nuclear Science & Technology)Sihver, L.*; Gustafsson, K.*; Mancusi, D.*; Sato, Tatsuhiko; Niita, Koji*
no journal, ,
Accurate particle and heavy ion transport codes are required for estimating the radiation risk for personnel and electronic equipment on spacecrafts, as well as estimating the effects of different shielding materials. We are therefore developing and benchmarking a general purpose Monte Carlo code PHITS, which can deal with the transports of all kinds of hadrons and heavy ions with energies up to 200 GeV/n. For the purpose of examining the applicability of PHITS to the shielding design, the absorbed doses in a water phantom inside an imaginary space vessel has been estimated for different shielding materials of different thicknesses. The results confirm previous results which indicate that PHITS is a suitable tool when performing shielding design studies of spacecrafts.
