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blanketYokoyama, Kenji; Shono, Akira*
International Handbook of Evaluated Reactor Physics Benchmark Experiments (CD-ROM), 223 Pages, 2010/03
Under the framework of the International Reactor Physic Experimental Evaluation Project (IRPhEP) organized by OECD/NEA, eight types of nuclear characteristics parameters, criticality, control rod worth, sodium void reactivity, fuel replacement reactivity, isothermal temperature coefficient and burnup reactivity coefficients, measured in the experimental fast reactor JOYO MK-I core were evaluated. In the present evaluation, not only nominal values but also uncertainties of experiments and analytical models were fully re-investigated according to the evaluation policy of the IRPhEP. In addition, each evaluation provides a reactor physics benchmark problem, which are expected to utilize for validating analytical models and nuclear data.
Yokoyama, Kenji; Shono, Akira*; Ishikawa, Makoto
Nuclear Science and Engineering, 157(3), p.249 - 263, 2007/11
Times Cited Count：5 Percentile：37.19(Nuclear Science & Technology)Experimental data acquired in the experimental fast reactor JOYO MK-I performance tests in the late 1970s have been revaluated and analyzed with a nuclear analysis system for fast reactors used in Japan Atomic Energy Agency (JAEA). For the purpose of improving the prediction accuracy of nuclear characteristics, nominal values and uncertainties of the experimental data were revaluated by using knowledge obtained after the MK-I performance test and calculation results based on the latest reactor physics analysis methods. All the nominal values were corrected by using a formulation of control rod interaction effects proposed in the present paper, and all the possible uncertainty factors were evaluated and quantified. The analysis results agreed well with measured values within the experimental and nuclear-induced uncertainties for all the nuclear characteristics of the criticality, control rod worth sodium void reactivity, fuel replacement reactivity and isothermal temperature coefficient.
Hazama, Taira; Shono, Akira*; Yokoyama, Kenji
Proceedings of American Nuclear Society Topical Meeting on Physics of Reactors (PHYSOR 2006) (CD-ROM), 10 Pages, 2006/09
Benchmark data have been created on reactor physics experiments performed in two reactors: the experimental fast reactor JOYO MK-I and Deuterium Critical Assembly (DCA). The data were prepared for the International Reactor Physics Experiment Evaluation Project (IRPhEP). In JOYO data, five kinds of reactivity data were evaluated: (1)criticality, (2)control rod worth, (3)sodium void reactivity, (4)fuel replacement reactivity, and (5)isothermal temperature coefficient. In particular, the control rod worth, a key quantity in all the reactivity evaluations, were evaluated in detail, considering interaction effects. In DCA data, three kinds of parameters were evaluated: (1)critical moderator level, (2)epithermal capture ratio of 
U, (3)dysprosium thermal reaction rate distribution in a fuel assembly. Data are systematically arranged in eight kinds of core configurations, varying the assembly pitch and void fraction. Each of evaluated data has a unique feature and will be useful to validate reactor physics calculation schemes.
Yokoyama, Kenji; Numata, Kazuyuki*; Shono, Akira; Ishikawa, Makoto
JNC TN9400 2005-024, 372 Pages, 2005/05
JNC-TN9400-2005-024.pdf:25.83MB
"JOYO" MK-I has a typical fast breeder reactor core which was fuelled by plutonium-uranium mixed oxide (MOX) and the fuel region was surrounded by blanket consisting of depleted uranium oxide. Since it has a simple core geometry without any other special irradiation subassembly, it is suitable for the reactor physics analysis. The experimental data acquired in the "JOYO" MK-I performance tests are analyzed with the latest analysis methods in order to resister the data in the IRPhE (International Reactor Physics Benchmark Experiments) project. For this analysis, nominal values and uncertainties of the experimental data were reevaluated by using knowledge obtained after the MK-I performance test and calculation results based on the latest analysis methods. As the nuclear characteristics for the analysis, we selected the criticality, the control rod worth, the isothermal temperature coefficient, the fuel replacement reactivity worth and the sodium void reactivity worth, which were measured in unburnt cores prior to the first power ascension to 50MWth. In this evaluation, not only the measurement uncertainty but also geometry uncertainty and composition uncertainty are considered. All the uncertainties are evaluated with a classification into the random and the systematic components, according as the guideline of IRPhE. The analysis of "JOYO" MK-I was previously carried out in the fiscal year of 1999. In this report, the analysis is totally carried out again with the revised group constants and the latest analysis codes such as the ultra-fine group lattice calculation code. In order for use in the benchmark problem preparation, the analytical results are shown with various correction factors that are estimated from the difference between analytical models. The uncertainties of analysis results are also evaluated by considering the correction factors. Furthermore, a consistency evaluation with the other existing critical experiment data is carried out by performing ...
Hazama, Taira; Iwai, Takehiko*; Shono, Akira
JNC TN9400 2005-011, 114 Pages, 2004/10
JNC-TN9400-2005-011.pdf:7.68MB
A program is in progress to dispose excess weapon plutonium in BN-600 fast reactor. To support the program, a series of critical experiments that simulates plutonium loading in BN-600 (BFS critical experiment and analysis) was carried out, and has been analyzed in an international collaboration with Russian Institute of Physics and Power Engineering. This report describes analysis results of the critical experiments on the last core configurations BFS-62-5 and BFS-66-1. In addition, already-reported analysis results on BFS-62-1 to -4 cores are updated in a unified method. BFS-62-5 and 66-1 have different features such as use of MOX fuel in the central area and having the sodium plenum region above the fuel region, when compared with BFS-62-1 to -4 mainly consisting of uranium fuel. Despite the differences, major nuclear parameters were successfully analyzed with similarly good accuracy. Even the sodium void reactivity, an important safety parameter and sensitive to the core configuration change, was analyzed within nuclear data uncertainty. These results will contribute to the improvement in reliability of core design in the dismantled plutonium disposition program of Russia and the FBR feasibility study of Japan.
Shono, Akira; Hazama, Taira; Ishikawa, Makoto; Manturov, G.*
Proceedings of International Conference on the Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004), 95315 Pages, 2004/00
The present paper provides evaluation results of predicted uncertainty on nuclear parameters on the BN-600 hybrid core, a feasible option for Russian surplus weapons plutonium disposition. Covariance of nuclear group constant, analysis error, and experimental error are considered to predict uncertainties of the hybrid core nuclear parameters by applying the nuclear group constant adjustment method. Analysis results of BFS mockup and other fast reactor core experiments were reflected in the evaluation.
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
Analysis ResultsShono, Akira; Mantourov, G.
Nuclear Science and Engineering, 144(3), p.211 - 218, 2003/07
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)None
Hazama, Taira; Sato, Wakaei*; Ishikawa, Makoto; Shono, Akira
JNC TN9400 2003-035, 44 Pages, 2003/05
JNC-TN9400-2003-035.pdf:1.07MB
Collaboration between Russian Institute of Physics and Power Engineering (IPPE) and Japan Nuclear Cycle Development Institute (JNC) named (Investigation of neutronic-physical characteristics and their change when introducing large quantity of neptunium (Np) at different BFS critical assemblies) is under progress. This is the first report of the collaboration to describe experimental information and JNC analysis results on BFS-67 critical experiment. In BFS-67 experiment, nuclear characteristics (criticality, control rod worth, sodium void reactivity, reaction ratio, etc) were measured in 4 different cores with various amounts of Np and location. JNC analysis was perfomed based on a JNC standard analysis scheme as in the analyses of BFS-62 critical experiments. (1)Sensitivity coefficients of Np capture cross section for the sodium void reactivity and control rod worth are large enough and comparable to those of U-238 and Pu-239. This indicates the experimental data can be used to improve design accuracy of Np loaded core. (2)C/E values for the criticality show high accuracy of 0.995 independent of core patterns, indicating accuracy of the calculation is high enough. (3)Calculated values for the sodium void reactivity agree with experimental values within 1cent and there is no influence of Np loading on calculation accuracy. (4)Calculated values for the control rod worth agree with experimental values within experimental errors for enriched B4C control rod. Those for naturaI B4C slightly overestimate. An influence of Np loading is not observed. (5)Calculated values for the reaction ratio agree with experimental values within 5% for fission reactions, whereas those for capture reactions show nearly 10% of differences. Positions of foils used in the measurement should be reflected.
Shono, Akira; Okawachi, Yasushi
Journal of Nuclear Science and Technology, 39(Suppl.2), p.493 - 496, 2002/08
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Shono, Akira; Sugino, Kazuteru; Hazama, Taira; Ishikawa, Makoto
7C-04, 0 Pages, 2002/00
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Shono, Akira; Iwai, Takehiko*
Journal of Nuclear Science and Technology, 39(Suppl.2), p.1085 - 1088, 2002/00
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Okawachi, Yasushi; Shono, Akira
JAERI-Conf 2001-006, p.121 - 124, 2001/03
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Shono, Akira; Tsunoda, Hirokazu; Takemura, Morio; Handa, Hiroyuki
PNC TN9410 95-171, 280 Pages, 1995/06
All experiments planned in the JASPER (Japanese-American Shielding Program for Experimental Research) project have been conducted from '85 to '92. Results obtained from the post-experimental analyses are described in the annual reports ('86
'94). This report is intended to review and summarize the enormous information of the annual reports. For the evaluation, the following topics were chosen. (1)Bulk Shielding attenuation characteristics and analysis accuracy (2)Design-dependent shielding characteristics and analysis accuracy (3)Evaluation of the shielding cross section libraries (4)Improvements in shielding analysis methods. Major conclusions are briefly as follows. (a)Both bulk shielding attenuation characteristics and streaming characteristics of configurations consisting of several materials including boron-carbide(B
C), graphite, stainless steel, sodium and etc. were clarified and the analysis accuracy confirmed. (b)JSDJ2 (based on the JENDL-2) was demonstrated to be a better cross section library for shielding analysis by reviewing experimental analyses results. (c)The shielding analysis system for fast reactors using a 2-dimensional discrete ordinates code as the standard has been improved and verified. (d)Useful experiences have been gained in verifying both a Monte Carlo code and a 3-dimensional discrete ordinates code, as well as in optimizing various parameters including those for mesh spacing techniques. Another objective of this report is to specify key information which will be useful to review JASPER. For this purpose, information on experiment and analysis is presented in a table format by each experimental item. All technical reports related to JASPER are listed.
Shono, Akira; ; ; Hoshi, Takashi*; Kusunoki, Hiroyuki*; Hamada, Masao*
PNC TY9471 94-002, 48 Pages, 1994/03
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Shono, Akira; ; ; Hoshi, Takashi*; Kusunoki, Hiroyuki*; Hamada, Masao*
PNC TY9471 94-001, 539 Pages, 1994/03
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Shono, Akira; *; *; *; Kitamura, Masaharu*
Nihon Genshiryoku Gakkai-Shi, 27(08), 94 Pages,
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