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Yokoyama, Kenji; Sugino, Kazuteru; Ishikawa, Makoto; Maruyama, Shuhei; Nagaya, Yasunobu; Numata, Kazuyuki*; Jin, Tomoyuki*
JAEA-Research 2018-011, 556 Pages, 2019/03
JAEA-Research-2018-011.pdf:19.53MB
JAEA-Research-2018-011-appendix1(DVD-ROM).zip:433.07MBJAEA-Research-2018-011-appendix2(DVD-ROM).zip:580.12MBJAEA-Research-2018-011-appendix3(DVD-ROM).zip:9.17MB
We have developed a new unified cross-section set ADJ2017, which is an improved version of the unified cross-section set ADJ2010 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses; the values are stored in the standard database for FBR core design via the cross-section adjustment methodology, which integrates with the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. The ADJ2017 is based on Japan's latest nuclear data library JENDL-4.0 as in the previous version of ADJ2010, and it incorporates more information on integral experimental data sets related to minor actinides (MAs) and degraded plutonium (Pu). In the creation of ADJ2010, a total of 643 integral experimental data sets were analyzed and evaluated, and 488 of the integral experimental data sets were finally selected to be used for the cross-section adjustment. In contrast, we have evaluated a total of 719 data sets, and eventually adopted 620 integral experimental data sets to create ADJ2017. ADJ2017 shows almost the same performance as ADJ2010 for the main neutronic characteristics of conventional sodium-cooled MOX-fuel fast reactors. In addition, for the neutronic characteristics related to MA and degraded Pu, ADJ2017 improves the C/E values of the integral experimental data sets, and reduces the uncertainty induced by the nuclear data. ADJ2017 is expected to be widely used in the analysis and design research of fast reactors. Moreover, it is expected that the integral experimental data sets used for ADJ2017 can be utilized as a standard database of FBR core design.
Yokoyama, Kenji; Kitada, Takanori*
Journal of Nuclear Science and Technology, 56(1), p.87 - 104, 2019/01
Times Cited Count：4 Percentile：41.24(Nuclear Science & Technology)no abstracts in English
Yokoyama, Kenji
Robutsuri No Kenkyu (Internet), (70), 12 Pages, 2018/04
no abstracts in English
Yokoyama, Kenji; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 53(10), p.1622 - 1638, 2016/10
Times Cited Count：10 Percentile：68.36(Nuclear Science & Technology)On the basis of the minimum variance approach, the unified formulation for three types of the cross-section adjustment methods has been derived in a straightforward way without assuming the normal distribution. These methods are intended to minimize the variances of the predicted target core parameters, the adjusted cross-section set, and the calculated integral experimental values. The first and the second methods are found to be slightly different from the extended and the conventional cross-section adjustment methods based on the Bayesian approach with the normal distribution assumption, respectively. However, they become equivalent in some cases and results. The third method is a new method, which is necessary from the viewpoint of the symmetry of the formulation. The derivation procedure proposed in the present paper is potentially applicable to developing more sophisticated cross-section adjustment methods because of the less assumptions on the probability density function.
Iwamoto, Hiroki; Nishihara, Kenji; Katano, Ryota*; Fukushima, Masahiro; Tsujimoto, Kazufumi
JAEA-Research 2014-033, 82 Pages, 2015/03
JAEA-Research-2014-033.pdf:6.53MB
The effect of experiments using Transmutation Physics Experimental Facility (TEF-P) is analysed from the viewpoint of the reduction of uncertainties in reactor physics parameters (criticality and coolant void reactivity) of an accelerator-driven system (ADS). The analysis is conducted by the nuclear-data adjustment method using JENDL-4.0 on the assumption that ve types of reactor physics experiments (a total of 44 experiments) are performed in TEF-P: (1) criticality experiment, (2) lead void reactivity experiment, (3) reaction rate ratio experiment, (4) sample reactivity experiment, and (5) fuel replacement reactivity experiment. As the result, 1.0% of uncertainty in criticality is found to be reduced to approximately 0.4%, and effective experiments for the reduction of uncertainty in criticality and coolant void reactivity are shown to be fuel replacement reactivity experiments and lead void reactivity experiments, respectively. Although these effects depend largely on the composition and amount of minor-actinide (MA) fuels, it is found that a combination of different types of experiments and database of existing experiments is effective in reducing the uncertainties.
Katakura, Junichi
JAERI-Conf 2001-009, 77 Pages, 2001/08
JAERI-Conf-2001-009.pdf:5.31MB
no abstracts in English
Sugino, Kazuteru; Iwai, Takehiko*;
JNC TN9400 2000-098, 182 Pages, 2000/07
JNC-TN9400-2000-098.pdf:5.74MB
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 ...
; Numata, Kazuyuki*; ; *; Oigawa, Hiroyuki*
JNC TY9400 2000-006, 162 Pages, 2000/04
JNC-TY9400-2000-006.pdf:4.57MB
no abstracts in English
Yokoyama, Kenji*; Numata, Kazuyuki*; Ishikawa, Makoto*; Oigawa, Hiroyuki; Iijima, Susumu
JNC-TY9400 2000-006, 168 Pages, 2000/04
no abstracts in English
; Numata, Kazuyuki*
JNC TN9400 2000-036, 138 Pages, 2000/03
JNC-TN9400-2000-036.pdf:10.16MB
Japan Nuclear Cycle Development lnstitute (JNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were renected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. ln this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor "JOYO" MK-l core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. 0n the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of "JOYO" MK-l core in comparison with ZPPR-9 core of JUPITER experiments.
*
JNC TJ9400 2000-005, 182 Pages, 2000/03
JNC-TJ9400-2000-005.pdf:4.74MB
The SLAROM code, performing fast reactor cell calculation based on a deterministic methodology, has been revised by adding the universal module PEACO of generating Ultra-fine group neutron spectra. The revised SLAROM, then, was utilized for evaluating reaction rate distributions in ZPPR-13A simulated by a 2-dim RZ homogeneous model, although actually ZPPR-13A composed of radial heterogereous cells. The reaction rate distributions of ZPPR-13A were also calculated by the code MVP, that is a continuous energy Monte Carlo calculation code based on a probabilistic methodology. By coparing both results, it was concluded that the module PEACO has excellent capability for evaluating highly accurate effective cross sections. Also it was proved that the use of a new fine group cross section library set (next generation set), reflecting behavior of cross sections of structural materials, such as Fe and O, in the fast neutron energy region, is indispensable for attaining a better agreement within 1% between both calculation methods. Also, for production of a next generation set of group cross sections, the code NJOY97.V107 was added to the group cross section production system and both front and end processing parts were prepared. This system was utilized to produce the new 70 group JFS-3 library using the evaluated nuclear data library JENDL-3.2. Furthermore, to confirm the capability of this new group cross section production system, the above new JFS-3 library was applied to core performance analysis of ZPPR-9 core with a 2-dim RZ homogeneous model and analysis of heterogeneous cells of ZPPR-9 core by using the deterministic method. Also the analysis using the code MVP was performed. Bycoaparison of both results the following conclusion has been derived; the deterministic method, with the PEACO module for resonance cross sections, contributes to improve accuracy of predicting reaction rate distributions and Na void reactivity in fast reactor cores. And it ...
Ando, Masaki; Iijima, Susumu; Ishikawa, Makoto*; Iwai, Takehiko*
JAERI-Tech 2000-025, p.45 - 0, 2000/03
JAERI-Tech-2000-025.pdf:1.77MB
no abstracts in English
Peter, J. Collins
PNC TN9410 97-034, 35 Pages, 1997/04
While in the Reactor Physics Research Section of the Advanced Technology Division at OEC, I participated in the project to construct a data library for the demonstration fast breeder reactor (DFBR). This library would be produced using a combination of evaluated differential cross sections together with integral experimental data for fast reactors, so as to assure sufficiently accurate calculations for the DFBR designs. I had much experience of the design and use of experiments for the large-size cores at ZPPR under the title JUPITER which was performed under the USDOE/PNC joint agreement. My contribution here was mainly in extension of the experimental database to include the very-hard spectrum fast criticals from the Los Alamos National Laboratory (LANL). The data for these cores are described. Our work at ANLW with the GMADJ code, which is similar in effect to the ABLE code that we use at PNC, showed why many experiments are important in this project as well as those in the more obvious Pu/U oxide conventional cores which are of current interest for the DFBR. This point was not appreciated at PNC and is discussed here. The data from the fast spectrum critical experiments made at Los Alamos are described together with information that I have been able to find concerning the uncertainties. The main interest is these experiments has been for prediction of criticality. Consequently, the full covariance information that we would like has not been published. However, the uncertainty in the fuel content is, by far, the major contributor to the uncertainty. The LANL experiments have been a principal leg of the data testing for fast reactors for all versions of ENDF/B in the US. For our work, they provide measurements at Mev energies which are not available from the experiments in the softer-spectrum of the LMFBR.
PNC TN9410 93-131, 139 Pages, 1993/05
In order to improve nuclear design method for large LMFBR cores, the cross-section adjustment method was fully applied to a 600MWe-class FBR core which was designed by the Plant Engineering office in FY1991. The features and performance or the new design method were compared with the conventional E/C bias correction method. The main results of the present study are: (a)From theoretical investigation for accuracy estimation formulas of various methods, the cross-section adjustment was found most effective to reflect integral information from critical experiments like JUPITER on FBR core design. (b)The design nominal values of the 600MWe-class FBR core were evaluated by various design methods. As a result, the design of FY1991 which used the bias method, was found reasonable to keep core performance and safety criteria, although some differences of the nominal values exist between the bias method and the adjustment method. (c)The design accuracy of various design methods was compared and the 1
- values were summarized in the table below. The cross-section adjustment method can be judged to possess sufficient applicability to large FBR core design and to improve prediction accuracy of nuclear core characteristics. ...
*; *; Sanda, Toshio*
PNC TJ9124 93-009, 334 Pages, 1993/03
To commertialize LMFBR, they are important subjects of research and development to improve the accuracy in neutronic design of large LMFBR cores and to make be able to design highly efficient core more rationaoy. The adjusted library has been made by being reflected the result of critical experiment of the JUPITER, etc. effectively as much as possible by using the cross-sections adjustment method based on the probabilistic theory. And the distinct improvement of the accuracy in neutronic design of large LMFBR cores has been achieved. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. Therefore the objective of the work is to improve the prediction accuracy for burnup characteristics using many burnup data of "Joyo" effectively. For the objective computer codes for analyzing sensitivity coefficients of burnup characteristics were prepared and cross-sections adjustment using burnup data of "Joyo" was done and the effect for the improvement of the accuracy for burnup characteristics was estimated. The results are as follows: (1)The computer codes which could analyze sensitivity coefficients of burnup characteristics of LMFBR taking into consideration plural cycles and refueling were prepared and their adequacy was made sure by comparing with direct calculations. (2)It was clarified that it is possible to improve the accuracy of burnup characteristics without affecting statically neutronic characteristics so much by applying the burnup characteristics to cross-sections adjustment.
Nakagawa, Tsuneo; Asami, Tetsuo
JAERI-M 87-025, 360 Pages, 1987/02
no abstracts in English
