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Barucci, M. A.*; Reess, J.-M.*; Bernardi, P.*; Doressoundiram, A.*; Fornasier, S.*; Le Du, M.*; Iwata, Takahiro*; Nakagawa, Hiromu*; Nakamura, Tomoki*; Andr
, Y.*; et al.
Earth, Planets and Space (Internet), 73(1), p.211_1 - 211_28, 2021/12
Times Cited Count：20 Percentile：82.04(Geosciences, Multidisciplinary)The MMX InfraRed Spectrometer (MIRS) is an imaging spectrometer on board of MMX JAXA mission. MIRS is built at LESIA-Paris Observatory in collaboration with four other French laboratories, collaboration and financial support of CNES and close collaboration with JAXA and MELCO. The instrument is designed to fully accomplish MMX's scientific and measurement objectives. MIRS will remotely provide near-infrared spectral maps of Phobos and Deimos containing compositional diagnostic spectral features that will be used to analyze the surface composition and to support the sampling site selection. MIRS will also study Mars atmosphere, in particular to spatial and temporal changes such as clouds, dust and water vapor.
Yokoyama, Kenji; Tatsumi, Masahiro*; Hirai, Yasushi*; Hyodo, Hideaki*; Numata, Kazuyuki*; Iwai, Takehiko*; Jin, Tomoyuki*; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; et al.
JAEA-Data/Code 2010-030, 148 Pages, 2011/03
JAEA-Data-Code-2010-030.pdf:3.23MB
A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional system), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system. On the other hand, burnup analysis functionality for power reactors as improved compared with the conventional system. In the development of MARBLE, the object oriented technology was adopted. As a result, MARBLE became an assembly of components for building an analysis code (i.e. framework) but not an independent analysis code system which simply receives input and returns output. Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system, SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS.
Hasegawa, Makoto; Kondo, Hitoshi; Kamei, Gento; Hirano, Fumio; Mihara, Morihiro; Takahashi, Kuniaki; Funabashi, Hideyuki; Kawatsuma, Shinji; Ueda, Hiroyoshi*; Oi, Takao*; et al.
JAEA-Research 2011-003, 47 Pages, 2011/02
JAEA-Research-2011-003.pdf:3.99MB
In 2009, NUMO and JAEA set up a technical commission to investigate the reasonable TRU waste disposal following a cooperation agreement between these two organizations. In this report, the calculation result of radionuclide transport for a TRU waste geological disposal system was described, by using the TIGER code and the GoldSim code at identical terms. Comparing the calculation result, a big difference was not seen. Therefore, the reliability of both codes was able to be confirmed. Moreover, the influence on the disposal site design (Disposal capacity: 19,000 m
) was examined when 10% of the amount of TRU waste increased. As a result, it was confirmed that the influence of the site design was very little based on the concept of the Second Progress Report on Research and Development for TRU Waste Disposal in Japan.
Yokoyama, Kenji; Hirai, Yasushi*; Hyodo, Hideaki*; Tatsumi, Masahiro*
JAEA-Data/Code 2009-016, 100 Pages, 2010/02
JAEA-Data-Code-2009-016.pdf:8.18MB
Development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the present study, we implemented functions for cell calculations and burnup calculations. With this, whole steps in analysis can be carried out with only this system. In addition, we modified the specification of user input to improve the convenience of this system. Since implementations being done so far had some bottlenecks to be resolved; we have realized the improvement on efficiency and amount of memory usage with modification on actual implementation.
Yokoyama, Kenji; Hirai, Yasushi*; Hyodo, Hideaki*; Tatsumi, Masahiro*
JAEA-Data/Code 2009-012, 208 Pages, 2010/02
JAEA-Data-Code-2009-012.pdf:11.28MB
Japan Atomic Energy Agency promotes development of innovative analysis methods and models in fundamental studies for next-generation nuclear reactor systems. In order to efficiently and effectively reflect the latest analysis methods and models to primary design of prototype reactor and/or in-core fuel management for power reactors, a next-generation analysis system MARBLE has been developed. In the present study, we examined in detail the existing design and implementation of ZPPR critical experiment analysis database followed by unification of models within the framework of the next-generation analysis system by extending to various critical experiment analysis. Furthermore, we examined requirements for functions of analysis results correction which is indispensable for critical analysis system, and designed and implemented an analysis system for various critical experiments including ZPPR.
Hirai, Yasushi*; Hyodo, Hideaki*; Tatsumi, Masahiro*; Yokoyama, Kenji
JAEA-Data/Code 2008-021, 110 Pages, 2008/10
JAEA-Data-Code-2008-021.pdf:3.47MB
Development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study on "Development of Burnup Analysis System for Fast Reactors (2)" in FY2006, design and implementation of models for detailed geometry of assembly, fuel loading pattern and so on, accompanied with specification and implementation of input file handling to construct data models. In this study, a prototype system has been implemented in which functionalities are embedded for calculation of macroscopic cross section, core calculation and burnup calculation applying the fruits of the study "Development of a Framework for the Neutronics Analysis System for Next Generation (2)". It also implements a fuel reloading/shuffling function controlled with simple description in user input for multi-cycle burnup analysis.
Hirai, Yasushi*; Hyodo, Hideaki*; Tatsumi, Masahiro*; Jin, Tomoyuki*; Yokoyama, Kenji
JAEA-Data/Code 2008-020, 188 Pages, 2008/10
JAEA-Data-Code-2008-020.pdf:15.06MB
Japan Atomic Energy Agency promotes development of innovative analysis methods and models in fundamental studies for next-generation nuclear reactor systems. In order to efficiently and effectively reflect the latest analysis methods and models to primary design of prototype reactor and/or in-core fuel management for power reactors, a next-generation analysis system MARBLE has been developed. In this study, detailed design of a framework, its implementation and tests are conducted so that a Python system layer can drive calculation codes written in C++ and/or Fortran. It is confirmed that various type of calculation codes such as diffusion, transport and burnup codes can be treated in the same manner on the platform for unified management system for calculation codes with a data exchange mechanism for abstracted data model between the Python and the calculation code layers.
Yokoyama, Kenji; Hirai, Yasushi*; Tatsumi, Masahiro*; Hyodo, Hideaki*; Chiba, Go; Hazama, Taira; Nagaya, Yasunobu; Ishikawa, Makoto
Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8 Pages, 2008/09
A development project of the next generation neutronics analysis code system, MARBLE, has been launched in JAEA. A software platform and common data models for fast reactor neutronics analysis were developed to realize the new system. At present, a fast reactor burnup calculation system, ORPHEUS, has been implemented in the MARBLE system. The new system reproduced benchmark results by the conventional code system and it reduced input data preparation works with the help of the capabilities supported by common data model packages. The new system was validated in an analysis of a burnup reactivity coefficient measured in the experimental fast reactor JOYO. These results show that MARBLE/ORPHEUS can be adopted as a new standard neutronics analysis system for fast reactors.
Tatsumi, Masahiro*; Hyodo, Hideaki*
JAEA-Review 2008-038, 95 Pages, 2008/08
JAEA-Review-2008-038.pdf:2.09MB
The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of an analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequences become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. Therefore, systemization of the burnup sensitivity analysis code has been done with an object-oriented scripting language. (NB: This document is a translation of JNC TJ9410 2004-002 published in February 2005.)
Hirai, Yasushi*; Hyodo, Hideaki*; Tatsumi, Masahiro*
JAEA-Data/Code 2007-019, 133 Pages, 2007/11
JAEA-Data-Code-2007-019.pdf:16.41MB
There is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility can contribute actual core design work and improvement of prediction accuracy. In the previous study on "Development of Burnup Analysis System (for Fast Reactors)" in FY2005, basic design was conducted to define each component in the system(input, solver, edit) and how to drive them. In this study, detailed design of the system and implementation of the I/O component were conducted according to the results in the basic design followed by proto-typing implementation.
Hyodo, Hideaki*; Tatsumi, Masahiro*
JAEA-Data/Code 2006-018, 120 Pages, 2006/08
JAEA-Data-Code-2006-018.pdf:8.42MB
In order to utilize the measured burnup data for improvement on accuracy in reactor core design, it is important to minimize the methodological errors to retrieve physical meanings from experimental data. The system for neutronics analyses that has been developed as the JUPITER standard analysis method assumes geometry of critical assembly, thus the system has not been maintained in functionalities for analysis of composition change of fuel materials. Therefore, there is a potential restriction for the purpose of detail analysis due to extreme inefficiency that comes from variety of limitation on its functionalities. It is not sufficient to follow a predefined analysis sequence in burnup analysis for reactor core; it is also needed to change analysis sequence to examine modeling error in analysis or to retrieve calculated values in an intermediate computation step for interpretation of physical meanings. Therefore it is not complete with a simple join of each function; it is needed to develop a new system for burnup analysis of reactor cores with flexibility on composition and decomposition of analysis components such as cell and core calculations. In this work, necessary conditions are examined for a new burnup analysis system targeted to actual reactor cores from the results of a research on the current working set in burnup analysis. With results in the research, a set of conceptual and fundamental designs were done.
Yokoyama, Kenji; Ishikawa, Makoto; Tatsumi, Masahiro*; Hyodo, Hideaki*
Proceedings of International Topical Meeting on Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications (M&C 2005) (CD-ROM), 12 Pages, 2005/09
In this work, a new burnup sensitivity analysis code, PSAGEP, has been developed with help from the object-oriented technique and the scripting language Python. It was confirmed they are powerful to support complex numerical calculation procedure for burnup sensitivity analysis. PSAGEP was reborn from the conventional hard-to-use code system as a user-friendly code system which can calculate the sensitivity coefficients of the nuclear characteristics considering burnup effect to cross-section changes based on the generalized perturbation theory.
Tatsumi, Masahiro*; Hyodo, Hideaki*
JNC TJ9410 2004-002, 98 Pages, 2005/02
JNC-TJ9410-2004-002.pdf:2.14MB
A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR core, however, it is important to accurately estimate not only neutronics characteristics but also burnup characteristics. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core.The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of an analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequences become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this purpose, systemization of the burnup sensitivity analysis code has been done with an object-oriented scripting language.In this study, an examination was conducted for the two-layer controlling model of the conventional system using Python, the object-oriented scripting language, With the result of the examination, a new analysis system for burnup sensitivity, PSAGEP (Python-wrapped SAGEP-burn), was implemented.
Tatsumi, Masahiro*; Hyodo, Hideaki*
JNC TJ9400 2003-012, 109 Pages, 2004/02
JNC-TJ9400-2003-012.pdf:3.56MB
In this study, points that should be improved in the current SAGEP-BURN were clarified through analysis of a computational process. Then a prototype of new SAGEP-BURN system was designed and implemented in Python. It is confirmed that the new system gives the identical results with that by the conventional system. For further implementation, analysis and design based on an object-oriented approach have been done.
Mori, Koichi*; Sekine, Norio*; Sato, Hajime*; Shikano, Naoto*; Shimayu, Daisuke*; Shiwaku, Hideaki; Hyodo, Kazuyuki*; Oka, Hiroshi*
Japanese Journal of Medical Physics, 22(1), p.13 - 19, 2002/03
no abstracts in English
Mori, Koichi*; Shikano, Naoto*; Shiwaku, Hideaki; Sato, Hajime*; Sekine, Norio*; Sato, M.*; Hyodo, Kazuyuki*; Ando, Masami*
Konika X-Rei Shashin Kenkyu, 51(3), p.101 - 105, 2000/05
no abstracts in English
Uyama, Chikao*; Takeda, Toru*; Toyofuku, Fukai*; Tokumori, Kenji*; Ando, Masami*; Hyodo, Kazuyuki*; Itai, Yuji*; Shiwaku, Hideaki; Nishimura, Katsuyuki*
Medical Applications of Synchrotron Radiation, p.149 - 157, 1998/00
no abstracts in English
Hirai, Yasushi*; Tatsumi, Masahiro*; Hyodo, Hideaki*; Yokoyama, Kenji; Ishikawa, Makoto
no journal, ,
no abstracts in English
Hyodo, Hideaki*; Tatsumi, Masahiro*; Yokoyama, Kenji; Ishikawa, Makoto
no journal, ,
no abstracts in English
Yokoyama, Kenji; Ishikawa, Makoto; Tatsumi, Masahiro*; Hyodo, Hideaki*
no journal, ,
no abstracts in English
Yokoyama, Kenji; Tatsumi, Masahiro*; Hirai, Yasushi*; Hyodo, Hideaki*; Numata, Kazuyuki*; Iwai, Takehiko*; Jin, Tomoyuki*; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; et al.
no journal, ,
no abstracts in English
