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Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Morota, Hidetsugu*; Nasif, H.*; Tanaka, Masanobu*; Polunovskiy, E.*; Loughlin, M.*
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03
Detailed nuclear analyses for the latest ITER NB system are required to ensure that NB design conforms to the nuclear regulations for the ITER building and licensing. A variety of nuclear analyses was started for ITER NB system including a tokamak building of 50m 
35m 20m and outside the building by using a Monte Carlo code MCNP in 2009. MCNP geometry input data were successfully produced from simplified NB CAD data with the improved GEOMIT code, which automatically converts CAD data to MCNP geometry input data. We have performed calculations of the effective dose rates during DT operation and after shutdown, and activation of the NB components, etc.
Nasif, H. R.*; Masuda, Fukuzo*; Morota, Hidetsugu*; Iida, Hiromasa*; Sato, Satoshi; Konno, Chikara
Nuclear Technology, 180(1), p.89 - 102, 2012/10
Times Cited Count：4 Percentile：24.98(Nuclear Science & Technology)GEOMIT is the CAD/MCNP conversion interface code. It is developed to automatically generate Monte Carlo geometrical data from CAD data due to the difference in the representation scheme. GEOMIT is capable of importing different CAD format as well as exporting different CAD format. GEOMIT has a capability to produce solid cells as well as void cells without using complement operator. While loading the CAD shapes (Solids), each shape is assigning material number and density according to its color on the original CAD data. Shape fixing process is been applied to cure the errors in the CAD data. Vertices location correctness is evaluated first, then a removal of free edges and removal of small faces processes. Binary Space Portioning (BSP) tree technique is used to automatically split complicated solids into simpler cells to avoid excessive complicated cells for MCNP to run faster. MCNP surfaces are subjected to an automatic reduction before creating the model. CAD data of International Thermonuclear Experimental Reactor (ITER) benchmark model has been converted successfully to MCNP geometrical input. MCNP input model validations have been carried out by checking lost particles and comparing volumes calculated by MCNP to those of the original CAD data. Different test cases have been evaluated for the ITER, include Blanket first wall heat loading calculations, surface fluxes and volume fluxes at different divertor regions as well as TF coils heating.
Sato, Satoshi; Nashif, H.*; Masuda, Fukuzo*; Morota, Hidetsugu*; Iida, Hiromasa*; Konno, Chikara
Fusion Engineering and Design, 85(7-9), p.1546 - 1550, 2010/12
Times Cited Count：9 Percentile：52.91(Nuclear Science & Technology)Automatic conversion systems from CAD data to MCNP geometry input data have been developed to convert the CAD data of the fusion reactor with very complicated structure. So far, three conversion systems (GEOMIT-1, ARCNCP and GEOMIT-2) have been developed. The void data can be created in these systems. GEOMIT-1 was developed in 2007, and a lot of manual shape splitting works for the CAD data were required to successfully convert the complicated geometry. ARCNCP was developed in 2008. The algorithm has been drastically improved on automatic creation of ambiguous surface in ARCNCP, and manual shape splitting works can be drastically reduced. The latest system, GEOMIT-2, does not require additional commercial software packages, though the previous systems require them. It has also functions of the CAD data healing and the automatic shape splitting. The geometrical errors of the CAD data can be automatically revised by the healing function, and the complicated geometries can be automatically split into the simple geometries by the shape splitting function. Any manual works are not required in GEOMIT-2. The latest system is very useful for nuclear analyses of fusion reactors.
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：45.28(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.
Shaaban, N.*; Masuda, Fukuzo*; Nasif, H.*; Yamada, Masao*; Sawamura, Hidenori*; Morota, Hidetsugu*; Sato, Satoshi; Iida, Hiromasa; Nishitani, Takeo
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 7 Pages, 2006/07
no abstracts in English
Oikawa, Tetsukuni; Muramatsu, Ken; Kasahara, Takeo*; Kawamata, Kazuhiko*; Morota, Hidetsugu*
Proceedings of 5th International Conference on Probabilistic Safety Assessment and Management (PSAM-5), p.2119 - 2125, 2000/00
no abstracts in English
Iida, Hiromasa; Sato, Satoshi; Konno, Chikara; Nishitani, Takeo; Nasif, H.*; Masuda, Fukuzo*; Yamada, Masao*; Morota, Hidetsugu*
no journal, ,
The ITER benchmark problem has been analysed with a CAD/MCNP automatic conversion cord "GEOMIT" which is beeing developed by JAEA-FNS in cooperation with CSD. The analysis includes (1) First wall neutron load, (2) Neutron fluxes in divertor cassette (3) Inboard TF coil nuclear heat and (4) Neutron flux distribution behind the port plug rear in a equitorial port. Generally very similar results have been obtained by GEOMIT with those of other ITER participating parties, suggesting that development of GEOMIT is progressing smoothly.
Morota, Hidetsugu*; Yamada, Masao*; Nasif, H.*; Masuda, Fukuzo*; Iida, Hiromasa; Sato, Satoshi; Nishitani, Takeo; Tamamizu, Shigeyuki*; Karaki, Junichi*
no journal, ,
Development of CAD preprocessing tool "CrtVoid" is reported. The code extracts surface element information and re-arrange them as input data for CAD/MCNP automatic conversion cord "GEOMIT". Reports includes detail of the development and test results of confirmation of its functions.
Nasif, H.*; Masuda, Fukuzo*; Noha, S.*; Yamada, Masao*; Morota, Hidetsugu*; Iida, Hiromasa; Sato, Satoshi; Konno, Chikara; Nishitani, Takeo
no journal, ,
GEOMIT is the CAD/MCNP conversion interface code. It is developed to automatically generate MCNP surface and cell cards from CAD data. GEOMIT uses sequential query language (SQL) to store the CAD data in the form of database tables to minimize the memory usage during the conversion processes. GEOMIT has a capability to produce void cells as well as solid cells without using complement operator. CAD data of International Thermonuclear Experimental Reactor (ITER) benchmark model has been converted successfully to MCNP geometrical input. The first wall neutron loading calculations agree very well with other countries results.
Nasif, H.*; Morota, Hidetsugu*; Sukegawa, Atsuhiko
no journal, ,
ATTILA is the 3D multi-group Sn particle transport code with arbitrary order anisotropic scattering. The transport equation solves in first order form using a tri-linear discontinuous spatial differencing on an arbitrary tetrahedral mesh. The key benefits are that as approximate accurate as Monte Carlo code, but mach faster and most useful visual and quantitative pre/post processing, etc. This presentation is to set out a radiation shielding analysis for the JT-60. ATTILA has shown its successful applicability with JT-60 machine disregarding its complex geometry for fully 360-degree model in the near future.
Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Iida, Hiromasa*; Morota, Hidetsugu*; Nasif, H.*; Tanaka, Masanobu*
no journal, ,
Shielding analyse were conducted on ITER NB by Monte Carlo calculation code MCNP5 and Fusion Evaluated Nuclear Data Library FENDL-2.1. Neutron and 
-ray dose rate distributions during operation are evaluated in L2 and L3 room. Also, these are evaluated outside the building. By using the direct 1 step Monte Carlo method (D1S-MCNP), decay -ray dose rate distributions are evaluated in L2 and L3 room about 10 days after operation. Detailed calculation results are presented in this conference.
