Sugino, Kazuteru; Takino, Kazuo
JAEA-Data/Code 2019-011, 110 Pages, 2020/01
A deterministic discrete ordinates method (SN method) transport calculation code for three-dimensional hexagonal geometry has been developed as the MINISTRI code (Ver. 7.0). MINISTRI is based on the triangle-mesh finite difference method, which can perform neutron transport calculations with high accuracy for cores of fast power reactors and assemblies of the Russian BFS critical facility. The present study has derived a proper scheme for remarkably improving the convergence of MINISTRI by investigating the issue of previous MINISTRI (Ver. 1.1), which sometimes plays a poor convergence performance in calculations for large-scale power reactor cores. The verification test of improved MINISTRI has been carried out for various cores by setting the reference result as the multi-group Monte-Carlo calculation with the same cross-sections as used in MINISTRI. As a result, it is found that the agreements are within 0.1% for eigenvalues and within 0.7% for power distributions. Thus, the satisfying accuracy of MINISTRI has been confirmed. In order to reduce the calculation time, the initial diffusion calculation scheme and the parallel processing have been implemented. As a result, the calculation time is reduced to the approximately one tenth compared with previous MINISTRI. Furthermore, adoption of the treatment of the anisotropic cell streaming effect, preparation of the perturbation calculation tool, implementation of the function for specification of the triangle-mesh-wise material and merging of the hexagonal-mesh calculation code MINIHEX have been carried out. Thus, the versatility of MINISTRI has been enhanced.
Sugino, Kazuteru; Numata, Kazuyuki*; Ishikawa, Makoto; Takeda, Toshikazu*
Annals of Nuclear Energy, 130, p.118 - 123, 2019/08
In MA sample irradiation test data calculations, the neutron fluence during irradiation period is generally scaled by using dosimetry data in order to improve calculation accuracy. In such a case, appropriate correction is required to burnup sensitivity coefficients obtained by the conventional generalized perturbation theory because some cancellations occur in the burnup sensitivity coefficients. Therefore, a new formula for the burnup sensitivity coefficient has been derived with the consideration of the neutron fluence scaling effect (NFS). In addition, the cross-section-induced uncertainty is evaluated by using the obtained burnup sensitivity coefficients and the covariance data based on the JENDL-4.0.
Yokoyama, Kenji; Sugino, Kazuteru; Ishikawa, Makoto; Maruyama, Shuhei; Nagaya, Yasunobu; Numata, Kazuyuki*; Jin, Tomoyuki*
JAEA-Research 2018-011, 556 Pages, 2019/03
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.
Takino, Kazuo; Sugino, Kazuteru; Yokoyama, Kenji; Jin, Tomoyuki*; Oki, Shigeo
Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1214 - 1220, 2018/04
Takeda, Toshikazu*; Yokoyama, Kenji; Sugino, Kazuteru
Annals of Nuclear Energy, 109, p.698 - 704, 2017/11
A new cross section adjustment method has been derived in which systematic errors in measured data and calculated results of neutronics characteristics are estimated and removed in the adjustment. Bias factors which are the ratio between measured data and calculated results are used to estimate systematic errors. The difference of the bias factors from unity is caused generally by systematic errors and stochastic errors. Therefore by determining whether the difference is within the total stochastic errors of measurements and calculations, systematic errors are estimated. Since stochastic errors are determined for individual confidence levels, systematic errors are also dependent to the confidence levels. The method has been applied to cross section adjustments using 589 measured data obtained from fast critical assemblies and fast reactors. The adjustments results are compared with those of the conventional adjustment method. Also the effect of the confidence level to the adjusted cross sections is discussed.
Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
Sugino, Kazuteru; Takeda, Toshikazu*
Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.573 - 581, 2015/09
Kugo, Teruhiko; Sugino, Kazuteru; Uematsu, Mari Mariannu; Numata, Kazuyuki*
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 12 Pages, 2014/09
The present paper summarizes calculation results for an international benchmark proposed under the framework of the Working Party on scientific issues of Reactor Systems (WPRS) of the Nuclear Energy Agency of the OECD. It focuses on the large size oxide-fueled SFR. Library effect for core performance characteristics and reactivity feedback coefficients is analyzed using sensitivity analysis. The effect of ultra-fine energy group calculation in effective cross section generation is also analyzed. The discrepancy is about 0.4% for a neutron multiplication factor by changing JENDL-4.0 with JEFF-3.1. That is about -0.1% by changing JENDL-4.0 with ENDF/B-VII.1. The main contributions to the discrepancy between JENDL-4.0 and ENDF/B-VII.1 are Pu capture, U inelastic scattering and Pu fission. Those to the discrepancy between JENDL-4.0 and JEFF-3.1 are Na inelastic scattering, Fe inelastic scattering, Pu fission, Pu capture, Pu fission, U inelastic scattering, Pu fission and Pu nu-value. As for the sodium void reactivity, JEFF-3.1 and ENDF/B-VII.1 underestimate by about 8% compared with JENDL-4.0. The main contributions to the discrepancy between JENDL-4.0 and ENDF/B-VII.1 are Na elastic scattering, Na inelastic scattering and Pu fission. That to the discrepancy between JENDL-4.0 and JEFF-3.1 is Na inelastic scattering. The ultra-fine energy group calculation increases the sodium void reactivity by 2%.
Buiron, L.*; Rimpault, G*; Fontaine, B.*; Kim, T. K.*; Stauff, N. E.*; Taiwo, T. A.*; Yamaji, Akifumi*; Gulliford, J.*; Fridmann, E.*; Pataki, I.*; et al.
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 16 Pages, 2014/09
Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD, an international collaboration is ongoing on the neutronic analyses of several Generation-IV Sodium-cooled Fast Reactor (SFR) concepts. This paper summarizes the results obtained by participants from institutions of different countries (ANL, CEA, ENEA, HZDR, JAEA, CER, KIT, UIUC) for the large core numerical benchmarks. These results have been obtained using different calculation methods and analysis tools to estimate the core reactivity and isotopic composition evolution, neutronic feedbacks and power distribution. For the different core concepts analyzed, a satisfactory agreement was obtained between participants despite the different calculation schemes used. A good agreement was generally obtained when comparing compositions after burnup, the delayed neutron fraction, the Doppler coefficient, and the sodium void worth. However, some noticeable discrepancies between the k-effective values were observed and are explained in this paper. These are mostly due to the different neutronic libraries employed (JEFF3.1, ENDFB7.0 or JENDL-4.0) and to a lesser extent the calculations methods.
Ohgama, Kazuya; Oki, Shigeo; Sugino, Kazuteru; Okubo, Tsutomu
Journal of Nuclear Science and Technology, 51(4), p.558 - 567, 2014/04
Core characteristics of a sodium-cooled fast breeder reactor (FBR) with 750 MWe output using highly decontaminated uranium and plutonium and highly minor-actinide-containing compositions were evaluated using the fast reactor cross-section set generated by the new Japanese nuclear data library JENDL-4.0. The core characteristics were compared with those obtained using the unified cross-section set ADJ2000R in order to investigate the differences between both the results. The effects on the core characteristics caused by the differences in the nuclear data of important reactions and nuclides in the cross-section sets were analyzed by a burnup sensitivity analysis. It was confirmed that adopting JENDL-4.0 to the FBR core design improves the breeding ratio, the burnup reactivity, and the reactivity control balance, because of the differences in the capture cross-sections of U-238 and Pu-239 of both the libraries. The difference in the sodium void reactivity evaluated with both the libraries was less than 1% because the increase caused by the differences in the elastic scattering cross-sections of sodium, the inelastic scattering cross-section, and the -average value of U-238 was practically cancelled out by the decrease caused by the differences in the capture cross-sections of Pu-239, the inelastic scattering cross-section of iron, and the capture cross- sections of Am-241.
Kawashima, Katsuyuki; Sugino, Kazuteru; Oki, Shigeo; Okubo, Tsutomu
Nuclear Technology, 185(3), p.270 - 280, 2014/03
Although the sodium void reactivity is limited up to 6 dollars in the current JSFR design, it should be significant to perform design studies of the low sodium void reactivity core besides the reference design, to increase the design margin considering any influence of the TRU fuel compositions. In this study, the BUMPY core is proposed as the low sodium void core concept, in which the partial-length fuels with upper sodium plenum are interspersed within the core, causing the steps in fuel length in the neighboring fuel assemblies. The void reactivity is considerably reduced due to the upward and lateral neutron leakage from the fuel region to the sodium plenum upon voiding. The BUMPY core is applied to the JSFR design. The calculated void reactivity of the BUMPY core is 2.5 dollars, which is considerably reduced from 5.3 dollars for the reference core. Moreover, the Doppler coefficient is almost the same as the reference core.
JAEA-Conf 2013-002, p.53 - 58, 2013/10
In order to contribute the validation of the cross-section covariance data, an equality was investigated between uncertainties of core characteristics evaluated by the conventional mock-up experimental approach and the current uncertainty quantification one.
Okumura, Keisuke; Sugino, Kazuteru; Kojima, Kensuke; Jin, Tomoyuki*; Okamoto, Tsutomu; Katakura, Junichi*
JAEA-Data/Code 2012-032, 148 Pages, 2013/03
A set of cross section libraries for the isotope generation and depletion calculation code ORIGEN2 was produced by using recent nuclear data JENDL-4.0. In this new library (ORLIBJ40), neutron-induced cross sections, fission product yields, isomeric ratios and half-lives were updated. ORLIBJ40 includes 24 libraries for typical UO or MOX fuels of PWR and BWR. In addition, it includes 36 libraries for various fast reactor fuels. ORLIBJ40 was applied to the post irradiation examination analyses of LWR nuclear spent fuels. As a result, it was confirmed that improvements were achieved especially for inventory and radioactivity estimations of minor actinides (Am and Cm isotopes) and fission products sensitive to cross sections (Eu and Sm isotopes) and for long-lived fission products (Se, etc.), compared with other existing ORIGEN2 libraries.
Sugino, Kazuteru; Numata, Kazuyuki*; Ishikawa, Makoto
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 9 Pages, 2013/03
Calculation accuracy of the sodium void reactivity for safety-enhanced fast reactor core concepts was evaluated with analyses of critical experiments. In these concepts, heterogeneous core configuration and sodium plenum replacement are adopted to reduce the sodium void reactivity to around zero. In the past, a variety of critical experiments for heterogeneous cores had been carried out in the ZPPR facility, some of which are compiled in the IRPhEP handbook. Further, several experiments for core with sodium plenum had been performed in the BFS-2 facility. Calculation analyses of above mentioned critical experiments have been performed by using the Japanese current reactor physics analytical system. These analyses clarified following items: (1) Accuracy for the axially-heterogeneous core was comparative or less to that of the homogeneous core. However, accuracy for the radially-heterogeneous core was not satisfactory. (2) Accuracy for the core with sodium plenum was not satisfactory in the sodium plenum voiding case.
Uematsu, Mari Mariannu; Sugino, Kazuteru; Kawashima, Katsuyuki; Okano, Yasushi; Yamaji, Akifumi; Naganuma, Masayuki; Oki, Shigeo; Okubo, Tsutomu; Ota, Hirokazu*; Ogata, Takanari*; et al.
JAEA-Research 2012-041, 126 Pages, 2013/02
The characteristics of sodium-cooled metal fuel core compared to MOX fuel core are given by its higher heavy metal density and superior neutron economy. By taking advantage of these characteristics and allowing flexibility in metal fuel specification and core design conditions as sodium void reactivity and bundle pressure drop, core design with high burnup, high breeding ratio and low fuel inventory features will be achievable. On ground of the major achievements in metal fuels utilization as driver fuels in sodium fast reactors in U.S., the metal fuel core concept is selected as a possible alternative of MOX fuel core concept in FaCT project. This report describes the following items as a result of the joint study on "Reactor core and fuel design of metal fuel core of sodium-cooled fast reactor" conducted by JAEA and CRIEPI during 4 years from fiscal year 2009 to 2012.
Nippon Genshiryoku Gakkai Dai-44-Kai Robutsuri Kaki Semina Tekisuto, p.165 - 180, 2012/08
no abstracts in English
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.
Iwamoto, Hiroki; Nishihara, Kenji; Tsujimoto, Kazufumi; Sugino, Kazuteru; Numata, Kazuyuki*
JAEA-Research 2011-036, 64 Pages, 2012/01
An analytical study of minor actinide (MA) transmutation systems was conducted using JENDL-4.0, with a comparison to JENDL-3.3 in terms of reactor physics parameters (criticality, void reactivity and the Doppler reactivity) and those uncertainties. As objects of the analyses, Accelerator driven system (ADS) and MA loaded fast reactor (FR) were assumed. It was found that there were considerable changes for both systems. As the results of the sensitivity and uncertainty analysis, we found that the difference of the parameters of ADS is due mainly to the inelastic scattering cross sections of lead isotopes and several reactions of Am. For FR, a large difference of the void reactivity uncertainty results primarily from the covariance data of the inelastic cross section of Na.
JAEA-Data/Code 2011-018, 125 Pages, 2012/01
Transport calculation codes for 3D hexagonal geometry have been developed: nodal code NSHEX, hexagonal-mesh finite difference code MINIHEX, and triangle-mesh finite difference code MINISTRI, which can perform neutron transport calculations with high accuracy for fast reactor cores. NSHEX is improved by extending the polynomial expansion order of intranode fluxes for better accuracy. MINIHEX is modified in the process of negative flux fix-up. In addition, MINISTRI is newly produced by changing the basic algorithm of MINIHEX. NSHEX, MINIHEX, and MINISTRI are applied to various fast reactor cores. It is found that these codes show satisfactory performance in terms of calculation accuracy. However, reduction of calculation time and improvement of convergence performance are required for all the codes by such measure as introduction of suitable acceleration technique. Further, development of post process functions is desired, which is represented by perturbation calculation scheme.
Sugino, Kazuteru; Jin, Tomoyuki*; Hazama, Taira; Numata, Kazuyuki*
JAEA-Data/Code 2011-017, 44 Pages, 2012/01
Fast reactor group constant sets UFLIB.J40 and JFS-3-J4.0 were prepared, which are based on the latest Japanese evaluated nuclear data library JENDL-4.0. Concerning UFLIB.J40, several fine group constant sets, which covered 70-group, 73-group, 175-group and 900-group structures, and the ultra fine group constant set were prepared. The number of nuclides for cross-sections of lumped fission products was extended so as to follow the extension of the number of fissile species for fission yield data.