Sugino, Kazuteru; Ishikawa, Makoto; Numata, Kazuyuki*; Iwai, Takehiko*; Jin, Tomoyuki*; Nagaya, Yasunobu; Hazama, Taira; Chiba, Go*; Yokoyama, Kenji; Kugo, Teruhiko
JAEA-Research 2012-013, 411 Pages, 2012/07
Aiming at evaluating the core design prediction accuracy of fast reactors, various kinds of fast reactor core experiments/tests have been analyzed with the Japan's latest evaluated nuclear data library JENDL-4.0. Totally 643 characteristics of reactor physics experiments/tests and irradiation tests performed using the critical facilities: ZPPR, FCA, ZEBRA, BFS, MASURCA, ultra-small cores of LANL and power plants: SEFOR, Joyo, Monju were dealt. In analyses, a standard scheme/method for fast reactor cores was applied including detailed or precise calculations for best estimation. In addition, results of analyses were investigated from the viewpoints of uncertainties caused by experiment/test, analytical modeling and cross-section data in order to synthetically evaluate the consistency among different cores and characteristics. Further, by utilizing these evaluations, prediction accuracy of core characteristics were evaluated for fast power reactor cores that are under designing in the fast reactor cycle technology development (FaCT) project.
Sugino, Kazuteru; Ishikawa, Makoto; Yokoyama, Kenji; Nagaya, Yasunobu; Chiba, Go; Hazama, Taira; Kugo, Teruhiko; Numata, Kazuyuki*; Iwai, Takehiko*; Jin, Tomoyuki*
Journal of the Korean Physical Society, 59(2), p.1357 - 1360, 2011/08
In order to improve the prediction accuracy of core performances in the fast reactor core design study, the unified cross-section set has been developed in Japan. The unified cross-section set, which combines a wide range of integral experimental information with differential nuclear data, is produced by using the cross-section adjustment technique based on the Bayesian parameter-estimation theory. A new set ADJ2010 is currently under development. The present paper describes the results of the cross-section adjustment for ADJ2010 which is based on the JENDL-4.0 data. The evaluation of the core design accuracy for a commercial power fast reactor core is also discussed. ADJ2010 will be released soon and will be expected to be utilized for core design of future fast reactors.
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
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.
Ozawa, Masaki; Iwai, Takehiko*; Babain, V.*; Shadrin, A.*
Proceedings of International Solvent Extraction Conference "Solvent Extraction-Fundamentals to Industrial Applications" (ISEC 2008), p.623 - 628, 2008/09
Bifunctional organophosphorus extractants dissolved in polar fluorinated diluents were studied aiming at directly recovering all f-elements from the dissolver solution of spent nuclear fuel. Distribution ratios of U, Np and Pu were sufficiently high for 0.40.8MCMPO in this solvent system, and combination of salt-free, methylamine carbonate (MAC), etc, were evaluated to obtain fractional stripping of f-elements. Static multi-stage extraction using artificial FBR dissolver solution supported the process feasibility.
Chiba, Go; Iwai, Takehiko*; Numata, Kazuyuki*; Hazama, Taira
JAEA-Research 2007-051, 52 Pages, 2007/07
A benchmark test of the latest evaluated nuclear data files, JENDL-3.3, JEFF-3.1 and ENDF/B-VII, has been carried out for fast reactor neutronics application. For this benchmark test, experimental data obtained at fast critical assemblies and fast power reactors are utilized. This benchmark test concludes that ENDF/B-VII predicts the neutronics characteristics of fast neutron systems better than other nuclear data files.
Sugino, Kazuteru; Iwai, Takehiko*
Proceedings of American Nuclear Society Topical Meeting on Physics of Reactors (PHYSOR 2006) (CD-ROM), 10 Pages, 2006/09
MONJU experimental data analysis was performed by using the detailed calculation scheme for fast reactor cores in Japan. Subsequently, feasibility of MONJU integral data was evaluated by cross-section adjustment technique for the use of FBR nuclear core design. It is concluded that MONJU integral data is quite valuable for building up the standard data base for large FBR nuclear core design. In addition, it is found that the application of the updated data base has a possibility to remarkably improve the prediction accuracy of neutronic parameters for MONJU.
Hazama, Taira; Iwai, Takehiko*; Shono, Akira
JNC-TN9400 2005-011, 114 Pages, 2004/10
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.
; Sato, Wakaei*; Hazama, Taira; Iwai, Takehiko*; Ishikawa, Makoto
JNC-TN9400 2003-074, 401 Pages, 2003/08
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
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, ...
; ; Iwai, Takehiko*; Numata, Kazuyuki*
JNC-TN9400 2002-008, 241 Pages, 2002/04
In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engneering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 and BFS-62-2 cores. The BFS-62-1 core models the present BN-600, and contains the enriched UO fuel surrounded by the UO blanket. The BFS-62-2 core has the same layout as the BFS-62-1 but the blanket region was replaced with stainless steel shield. For core parameter analyses, the 3-D Hexagonal-Z or XYZ geometry model was applied by not only diffusion calculation but also transport calculation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality, the reaction rate ratio and reaction rate distribution in BFS-62-1. In the reaction rate distribution of BFS-62-2 calculation without cross-section adjustment produced big radial dependency of calculation over experiment value (C/E value) in the core region and overestimation in the shield region. Cross-section adjustment technique procedure improved those estimation, however alternation of cross-section of Iron, which was dominant in above improvement, compared to the cross-section error, and further investigation was required. Concerning the control rod worth of BFS-62-1, radial dependency of the C/E value was observed whether cross-section adjustment technique was applied or not, therefore comparison with results of other BFS-62 ...
; Iwai, Takehiko*
Journal of Nuclear Science and Technology, 39(Suppl.2), p.1085 - 1088, 2002/00
; Iwai, Takehiko*;
JNC-TN9400 2000-098, 182 Pages, 2000/07
In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 assembly, which is the first core of the BFS-62 series. The core contains the enriched U0 fuel surrounded by the U0 blanket. The standard analytical method for fast reactors has been applied, which was used for the JUPITER and other experimental analyses. Due to the lack of the analytical data the 2D RZ core calculation was mainly used. The 3D XYZ core calculation was applied only for the preliminary evaluation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality and the reaction rate ratio. However, it was found that accurate evaluation of the reaction rate distribution was impossible without exact consideration of the arrangement of the two types of sodium (with and without hydrogen impurity), which can be accommodated by the 3D core analysis, thus it was essentia1. In addition, it was clarifie that there was a room for an improvement of the result on the reaction rate distribution in the blanket and shielding regions. The application of the 3D core calculation improved the result on the control rod worth because 3D core model can more exactly consider the shape of the control rod. Furthermore it was judged that the result of the analysis on the sodium void reactivity .....
; 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 ...
Oki, Shigeo; Iwai, Takehiko*; Jin, Tomoyuki*
JNC-TN9400 2000-080, 532 Pages, 2000/03
Transmutation Property of minor actinide nuclides (MA) was analyzed for fast reactor cores having different coolant and fuel types in order to obtain basic data for evaluating an ability of the efficient use of resource and reducing the effect on the environment. The investigated fast reactor cores were (a) sodium cooled and oxide fueled core, (b) lead cooled and nitride fueled core (BREST-300), (c) carbon dioxide gas cooled and oxide fueled core (ETGBR), (d)lead cooled and oxide fueled core, (e) sodium cooled and nitride fueled core (both He-bond and sodium-bond), and (f) sodium cooled and metallic fueled core. Followings were observed for the relation between MA transmutation property and the different types of coolant and fuel of fast reactor core. (1) For the MA transmutation rate, the relation "Oxide Metal Nitride" was found out for difference of fuel type. A main reason of the increment of MA transmutation rate is that the neutron flux level rises on nitride and metallic fueled cores in comparison with oxide core. (2) The relation "Lead Sodium and Carbon dioxide" can be seen for the MA transmutation rate in the difference of coolant, but it is not clear whether the cause is driven from the difference of coolant itself on the difference of core design. (3) The changes of MA transmutation property mentioned above are comparatively small.
Ando, Masaki; Iijima, Susumu; Ishikawa, Makoto*; Iwai, Takehiko*
JAERI-Tech 2000-025, p.45 - 0, 2000/03
no abstracts in English
Iwai, Takehiko*; Sugino, Kazuteru; Ishikawa, Makoto
PNC-TN9410 98-079, 54 Pages, 1998/07
In the Reactor Physics Research Section, O-arai Engineering Center, a Standard Data Base for Nuclear Design is under development to improve the accuracy of FBR nuclear design calculations. The Reactor Integral Test Working Group of the Sigma committee is compiling data from TCA(Tank Type Critical Assembly), STACY(Static Experimental Critical Facility) and FCA(Fast Critical Assembly), for a set of nuclear data benchmark problems. An FBR benchmark problem from the Working Group has been added to the Data Base: the ZPPR-9 core, simplest of the JUPITER-I series, recommended for its ease of use. The compiled nuclear parameters are Criticality; Reaction rate data(Reaction rate ratio & Reaction rate distribution); Control rod worth; Sodium void reactivity and Sample Doppler reactivity. The benchmark problem definition is an idealization of the experimental geometly; we used detailed analytical methods to prepare correction factors, so that users of the benchmark can compare their results with the experiment. Wide use of the benchmark problem is anticipated. However, the calculation of correction factors is affected by the mesh size, the number of energy groups and the definition of cross sections, so it is necessary to use detailed analytical methods to produce modified correction factors when one uses a calculational model in which mesh size etc. are different.
Sugino, Kazuteru; Iwai, Takehiko*
PNC-TN9410 98-067, 57 Pages, 1998/07
A standard data base for FBR core nuclear design is under development in order to improve the accuracy of FBR design calculation. As a part of the development, we investigated an improved treatment of double-heterogeneity and a method to calculate homogenized control rod cross sections in a commercial reactor geometry, for the betterment of the analytical accuracy of commercial FBR core characteristics. As an improvement in the treatment of double-heterogeneity, we derived a new method (the direct method) and compared both this and existing methods with continuous energy Monte-Carlo calculations. In addition, we investigated the applicability of the reaction rate ratio preservation method as a advanced method to calculate homogenized control rod cross sections. The present studies gave the following information: (1)An improved treatment of double-heterogeneity: for criticality the existing method showed good agreement with Monte-Carlo result within one standard deviation; the direct method was consistent with existing one. Preliminary evaluation of effects in core characteristics other than criticality showed that the effect of sodium void reactivity (coolant reactivity) due to the double-heterogeneity was large. (2)An advanced method to calculate homogenized control rod cross sections: for control rod worths the reaction rate ratio preservation method agreed with those produced by the calculations with the control rod heterogeneity included in the core geometry; in Monju control rod worth analysis, the presented method overestimated control rod worths by 1 to 2%compared with the standard method, but these differences were caused by more accurate model in the presented method and it is considered that this method is more reliable than the standard one. These two methods investigated in this study can be directly applied to core characteristics other than criticality or control rod worth. Thus it is concluded that these methods will contribute to the ...
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 ...
Iwai, Makoto; *; *; *; Shimizu, Takehiko*; Hayashi, Naomi*; *; Maruo, Yoshihiro
PNC-TN852 84-06, 1 Pages, 1984/03
This manual includes the standard procedures for analysis of radioactive materials and chemical pollutants in liquid and gaseous waste effluent discharged into the environment from the nuclear facilities of Tokai Works of PNC. The second edition, PNCT852 - 79 - 10 was published in 1979. Almost all analytical procedures have been modified and new technique has been used for these five years, so the third edition revised entirely was to be published this time. The principal consideration was placed on the revision of sample preparation and safety handling of the chemical agents in particular.