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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：13 Percentile：80.63(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.
Kudo, Atsunari; Kurabayashi, Kazuaki; Yanagibashi, Futoshi; Sasaki, Shunichi; Sato, Takehiko; Fujimoto, Ikuo; Obu, Tomoyuki
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
The Co-processing process is the extraction process to recover Pu/U mixed product solution with given Pu/U ratio for improving of nuclear proliferation resistance. In addition, Np is also recovered with U and Pu because Np is one of minor actinides and a long-lived radionuclide and Np has the extractability into TBP solvent. Development of its flowsheet achieves to decrease environmental effect of waste materials. The orientation of development about Co-processing process is to demonstrate of reprocessing the future spent fuels from a LWR, a LWR-MOX hybrid, and a FR-MOX with one cycle. We demonstrated by use of miniature reflux-type centrifugal contactors at the partitioning unit. The test conditions of the Pu/U ratio in the loaded solvents were 1%, 3%, and 5% considering the composition of spent fuels. We used the HAN as the reductant of Np (VI) for back extraction. The results of these tests were very good. We got the prospect of U, Pu, and Np Co-processing flowsheet.
Sako, Hiroyuki; Harada, Hiroyuki; Sakaguchi, Takao*; Chujo, Tatsuya*; Esumi, Shinichi*; Gunji, Taku*; Hasegawa, Shoichi; Hwang, S.; Ichikawa, Yudai; Imai, Kenichi; et al.
Nuclear Physics A, 956, p.850 - 853, 2016/12
Times Cited Count：12 Percentile：65.66(Physics, Nuclear)Uchida, Shunsuke; Hanawa, Satoshi; Nishiyama, Yutaka; Nakamura, Takehiko; Sato, Tomonori; Tsukada, Takashi; Kysela, J.*
Nuclear Technology, 183(1), p.119 - 135, 2013/07
Times Cited Count：5 Percentile：38.62(Nuclear Science & Technology)In-pile loop experiments are one of the key technologies which can provide an understanding of corrosion behaviors of structural materials in nuclear power plants (NPPs). The experiments should be supported not only by reliable measurement tools to confirm corrosive conditions under neutron and 
ray irradiations but also by theoretical models for extrapolating the measured data to predict corrosion behaviors in NPPs. The relationships among electrochemical corrosion potential (ECP), metal surface conditions, exposure time and other environmental conditions have been determined from in situ measurements of corrosion behaviors of stainless steel specimens exposed to H
O and O in high temperature water. Based on the relationships, a model to evaluate ECP of stainless steel was developed by coupling an electrochemical model and a double oxide layer model. Major conclusions obtained from the evaluation model are as follows. (1) The difference in ECP behaviors of the specimens exposed to HO and O were mainly from the thickness and developing rate of the inner oxide layers. (2) Calculated ECP behaviors, e.g., the different responses to HO and O and hysteresis and memory effects, agreed with the measured ones. (3) Neutron exposure might decrease ECP due to radiation-induced diffusion in the oxide layer. The ECP evaluation model will be applied to evaluation of corrosive conditions in the JMTR in-pile loop.
Yamada, Junya; Seya, Natsumi; Haba, Risa; Muto, Yasunobu; Numari, Hideyuki*; Sato, Naomitsu*; Nemoto, Koji*; Takasaki, Hiroichi*; Shimizu, Takehiko; Takasaki, Koji
JAEA-Data/Code 2013-006, 100 Pages, 2013/06
JAEA-Data-Code-2013-006.pdf:12.04MB
This report presents the results of emergency radiation monitoring, including ambient -ray dose rate, atmospheric radioactivity, meteorological observation and estimation of internal exposure resulting from the accident at the Fukushima Daiichi Nuclear Power Plant triggered by the earthquake off the pacific coast of Tohoku on 11th March 2011, conducted by Oarai Research and Development Center (ORDC), Japan Atomic Energy Agency (JAEA) from March to May, 2011. ORDC is located in the central part of Ibaraki prefecture and approximately 130 km southwest of the Fukushima Daiichi Nuclear Power Plant. From around 15th to 21st March, 2011, the ambient -ray dose rate increased and many radioactive nuclides were detected in the atmosphere.
Hosotani, Risa; Sato, Naomitsu*; Shimizu, Takehiko; KOBAYASHI, Hideo
Saikuru Kiko Giho, (25), p.25 - 32, 2004/00
To observe the airborne gamma radiation dose rate, monitoring posts are set up to a border of supervised area of JNC-OEC. Measurement values of some ionization chambers set at monitoring posts were increased by unknown origin signal at random times. To probe the cause, measurement of electric field intensity at the ionization chamber and immunity test at anechoic chamber are carry out. Result of examination made clear that measurement values are increased by specific frequency band electromagnetic wave. They were also clearly that ferrite cores and shield tube are effective as eliminate of an electromagnetic wave noise. When ferrite cores are attached to cables of ionization chambers, unknown increase of measurement value doesn't occur.
Shono, Akira; Sato, Wakaei*; Hazama, Taira; Iwai, Takehiko*; Ishikawa, Makoto
JNC TN9400 2003-074, 401 Pages, 2003/08
JNC-TN9400-2003-074.pdf:48.95MB
Nuclear design accuracy on the BN-600 hybrid core has been evaluated using the JNC's nuclear analysis system for FBR cores, by utilizing the critical experiment analysis results on BFS-62 configuration that had been obtained under JNC's efforts for Russian surplus weapons plutonium disposition. In the BN-600 hybrid core design, a part of the current UO2 fuel region is replaced by MOX fue1, and the Peripheral blanket region by stainless steel reflectors, respectively. These changes were simulated in a series of critical experiment configurations (BFS-62-1 to 4). Based on the analysis results on both BFS-62 configurations and other fast reactor cores, nuclear design accuracy on the BN-600 hybrid core has been evaluated by applying both the group constant adjustment method and the bias method. Evaluated nuclear parameters include, the criticality, fission rate distribution, sodium void reactivity, control rod worth, burn-up reactivity loss, etc. It is concluded, by applying the group constant adjustment method, that the evaluated accuracy (uncertainty) of most of the nuclear parameters can be decreased to less than half of those based on the basic nuclear constant without reflecting any experimental data. The improvement was mainly achieved by reducing the covariance of the iron elastjc cross section. This significant effect results from the feature of the BN-600 hybrid core, which has relatively larger power density, adopts U235 as the main fissile nucljde, and has the stainless steel reflector surrounding the fuel region. In addition, good consistency of analysis results between the BFS and other fast reactor cores is confirmed. Information obtained by BFS-62 experiment show significant contribution to the accuracy improvement. It is also found that the bias method shows less significant effects on the accuracy improvement than the group constant adjustment method. Furthermore, the bias method may degrade the accuracy for certain nuclear parameters that have large e
Hazama, Taira; ; Iwai, Takehiko*; Sato, Wakaei*
JNC TN9400 2002-036, 113 Pages, 2002/06
JNC-TN9400-2002-036.pdf:4.44MB
In order to support the Russian excess weapons plutonium disposition program, the intemational collaboration has started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering(IPPE). In the frame of the collaboration, analyses have been carried out for a series of critical experiments that simulate BN-600 (Russian commercial fast reactor). This report summarizes analysis results of the critical expeliments on BFS-62-3A and BFS-62-4 cores. BFS-62-3A core models BN-600 hybrid core in which the present BN-600 core is modified so as to partially load MOX fuel assemblies and replace the blanket region with stainless steel. BFS-62-4 core has the same layout as BFS-62-3A core except that the blanket region is not replaced. The analyses were performed with JNC standard method developed in the analysis of JUPITER experiment. The results show a good agreement with experimental values for the criticality and the reaction rate ratio. For the control rod worth and the reaction rate distribution, the results for BFS-62-4 core are also reasonable. However, for BFS-62-3A, analysis results overestimate the reaction rate in the stainless steel region by 20% and underestimate reactivity worth for one of the control rods by 10%. For the sodium void reactivity, underestimation of more than 20% were observed, but the disagreement were successfully solved by adopting a newly developed nuclear constant set with a fine group structure. In addition, analysis accuracies were compared among a series of analyses and it was confirmed that the introduction of MOX fuel assemblies does not affect the accuracy. The final goal of the work is to reflect the analysis results for designing BN-600 hybrid core. Then similarity was investigated between BFS-62-3A core and BN-600 hybrid core. A good similarity was found in the neutron spectrum, the fission reaction ratio, the fission reaction distribution, and the control rod worth. However, ...
Maki, Akira; ; Taguchi, Katsuya; ; Shimizu, Ryo; Shoji, Kenji;
JNC TN8410 2001-012, 185 Pages, 2001/04
JNC-TN8410-2001-012.pdf:9.61MB
"The third technological meeting of Tokai Reprocessing plant (TRP)" was held in JNFL Rokkasyo site on March 14
, 2001. The technical meetings have been held in the past two times. The first one was about the present status and future plan of the TRP and second one was about safety evaluation work on the TRP. At this time, the meeting focussed on the corrosion experrience, in-service inspection technology and future maintenance plan. The report contains the proceedings, transparancies and questionnaires of the meeting are contained.
; Sato, Wakaei*; Iwai, Takehiko*
JNC TN9400 2000-096, 113 Pages, 2000/06
This report describes the updated analyses results on the BFS-58-1-I1 core. The experiment was conducted at BFS-2 of Russian Institute of Physics & Power Engineering (IPPE). The central region is "non-Uranium fuel zone", where only Pu can induce fission reaction. The non-U zone is surrounded by MOx fuel zone, which is surrounded by U0 fuel zone. Sodium is used for simulating the coolant material. As it was found that the lattice pitch had been incorrectly understood in the past analyses, all items have been re-calculated using the corrected number densities. Furthermore, significantly softened neutron spectrum in the central region caused problems in applying the plate-stretch model that has been established for fast reactor cores through JUPITER experimental analyses. Both keeping the pellet density and using SRAC library for the elastic cross section for lighter nuclides allow us to obtain reasonable analysis accuracy on the spectral indices that were measured at the center of the core. Application of such a cell model was justified through comparison among various cell models using continuous energy Monte-Carlo code MVP. It is confirmed that both the MOX zone and the U0 zone can be correctly evaluated by the plate-stretch model. Based on the updated cell calculation, both the effective multiplication factor (k-eff)and the spectral indexes agree well with the measured values. The transport and mesh-size correction is made for the k-eff evaluation. Those results also agree well within reasonable difference between those obtained by IPPE and CEA, which were obtained by using sub-group method or continuous-energy Monte Carlo code. Evaluation by the nuclear data library adjustment confirmed that the analyses results of the BFS-58-1-I1 core have no significant inconsistency with JUPITER experimental analyses results. Those results are quite important for starting BFS-62 cores, which will be analyzed in the framework of supporting program for Russian ...
Ishikawa, Makoto; Sato, Wakaei*; Sugino, Kazuteru; Yokoyama, Kenji; Numata, Kazuyuki*; Iwai, Takehiko*
PNC TN9410 97-099, 512 Pages, 1997/11
A Standard data base for LMFBR core nuclear design has been developed to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as demonstration or commercial FBRs. To develop the data base, extensive work has been prerformed to accumulate and evaluate many kinds of results from fast reactor physics experiments and their analyses. The present report summarizes the analytical results of the JUPITER experiments, using the most recent nuclear data library (JENDEL-3.2) and the lates analytical methods in a consistent manner. In the present work, a great emphasis was placed on guaranteeing the essential requirements for this kind of general data base, that is, "Accountability", "Traceability" and "Consistency". In other words, consistent strategies and analytical methods were applied to all calculations, including detialed corrections; the enormous analytical input data generated were all saved in the form of computer files, so that reanalysis of any experiment could be easily performed for verification or in response to future improvement in nuclear data or analytical methods. The main results of the present JUPITER analysis are as follows: (1) The C/E(calculation/experiment) values of criticality were slightly underestimated by -0.7
-0.3%
k. (2) The reaction rate ratio of C28/F49 was overestimated by +4+6% with the standard analytical method. However it was found to improve about 2% after the cell factors were revised using the Monte Carlo method. (3) The radial space-dependency of the reaction rate distribution and control rod worth almost disappeared in the homogeneous cores. (4) The previous overestimation of sodium void reactivity was greatly improved in the homogeneous cores.
Sugino, Kazuteru; Yokoyama, Kenji; Ishikawa, Makoto; Sato, Wakaei*; Numata, Kazuyuki*; Iwai, Takehiko*
PNC TN9410 97-098, 247 Pages, 1997/11
The present report compiles the advances in experiment analyses of JUPITER, which was joint research programs between U.S.DOE and PNC of Japan, using the Zero Power Physics Reactor (ZPPR) large fast critical facility at ANL-Idaho in 1978 to l988. The advances here are use of the latest nuclear data library and the application of analytical methods which treat mechanisms in more detail or use fewer modeling approximations. As a result of using the latest nuclear data library, C/E values of nearly all characteristics approached unity, and the discrepancies between cores were reduced. Thus it is shown that the latest data library is effective for an analysis of nuclear characteristics. Further, an advance in analytical methods brought C/E value close to unity, and it clarifies the causes of differences between the calculational and experimental values. The current evaluation for each nuclear palameter shows following: (1)Criticality. The C/E values are from 0.993 to 0.997, a systematic underestimate. This underestimation is much smaller than the error caused by the uncertainty in nuclear data, which is the dominant error for this characteristic. In terms of analytical method, there are significant differences in calculation results between present and Monte-Carlo based methods, so more investigation will be required in future. (2)Doppler reactivity. The C/E values are from 0.8 to 0.9, a systematic underestimate. The analytical method, which is stood for by the use of ultra fine energy structure analysis, is so detailed that there is little room for improvement in that term. Therefore, some evaluation of the self-shielding factors and comparison with other Doppler reactivity experiments will be required. (3)Reaction rate distribution. It is judged that the present analytical method has an adequate accuracy for the core regions of homogeneous and axially heterogeneous cores, because the C/E values varied from unity by less than 2% for Pu-239 fission, U-235 fission ...
C at FCA, 1; Development of experimental device for Doppler reactivity worth measurement with small sample heated up to 1500COigawa, Hiroyuki; Okajima, Shigeaki; Mukaiyama, Takehiko; ; Hishida, Makoto; *; *; Kasahara, Y.*
JAERI-M 94-043, 46 Pages, 1994/03
no abstracts in English
U Doppler reactivity worth in FCA XVI-1 and XVI-2 coresOigawa, Hiroyuki; Okajima, Shigeaki; Mukaiyama, Takehiko;
JAERI-M 92-113, 36 Pages, 1992/08
no abstracts in English
; ; *; ; ; ; ; Ono, Akio; *; ; et al.
JAERI-M 85-045, 136 Pages, 1985/04
no abstracts in English
; ; ; ; ; ; ; ; Suzuki, Katsuo;
JAERI-M 82-103, 137 Pages, 1982/09
no abstracts in English
; ; ; ; ; ; ; ; ; Suzuki, Katsuo; et al.
JAERI-M 82-102, 368 Pages, 1982/09
no abstracts in English
; ; ; ; ; ; ; Suzuki, Katsuo; ;
JAERI-M 8399, 253 Pages, 1979/08
no abstracts in English
; ; ; ; ; ; ; ; ; Suzuki, Katsuo; et al.
JAERI-M 8064, 255 Pages, 1979/03
no abstracts in English
