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.
Yokoyama, Kenji; Maruyama, Shuhei; Numata, Kazuyuki; Ishikawa, Makoto; Takeda, Toshikazu*
Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.1906 - 1915, 2016/05
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%.
Uematsu, Mari Mariannu; Kugo, Teruhiko; Numata, Kazuyuki*
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 15 Pages, 2014/09
In the frame work of the working party on reactor and system (WPRS) of the OECD/NEA, the benchmark on SFR was conducted. Within the OECD/NEA/WPRS benchmark, study on medium size metallic fuel core was performed using a code system for fast reactor core calculation with deterministic method MARBLE and with a Monte Carlo method MVP. The latest nuclear library JENDL-4.0 is used for evaluation of eigenvalues (k) and reactivity (sodium void, Doppler and control rod worth) calculations. Depletion calculations are conducted using MARBLE/BURNUP with deterministic method for flux calculation and MVP-BURN with Monte Carlo method. The analysis results and discrepancies between different analysis methods are summarized in this paper. Sensibility studies of eigenvalue and sodium void reactivity of the medium size metallic fuel benchmark core are also conducted to determine the main reactions contributing to the difference between JENDL-4.0 and other libraries JEFF-3.1 and ENDF/B-VII.
Yokoyama, Kenji; Hazama, Taira; Numata, Kazuyuki*; Jin, Tomoyuki*
Annals of Nuclear Energy, 66, p.51 - 60, 2014/04
A comprehensive and versatile reactor analysis code system, MARBLE, has been developed. MARBLE is designed as a software development frame-work for reactor analysis, which offers reusable and extendible functions and data models based on physical concepts, rather than a reactor analysis code system. From a viewpoint of the code system, it provides a set of functionalities utilized in a detailed reactor analysis scheme for fast criticality assemblies and power reactors, and nuclear data related uncertainty quantication such as cross-section adjustment. MARBLE covers all phases required in fast reactor core design prediction and improvement procedures, i.e. integral experiment database management, nuclear data processing, fast criticality assembly analysis, uncertainty quantication, and power reactor analysis. In the present paper, these functionalities are summarized and system validation results are described.
Fukushima, Masahiro; Ishikawa, Makoto; Numata, Kazuyuki*; Jin, Tomoyuki*; Kugo, Teruhiko
Nuclear Data Sheets, 118, p.405 - 409, 2014/04
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.
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.
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.
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.
Matsui, Yoshinori; Takahashi, Hiroyuki; Yamamoto, Masaya; Nakata, Masahito; Yoshitake, Tsunemitsu; Abe, Kazuyuki; Yoshikawa, Katsunori; Iwamatsu, Shigemi; Ishikawa, Kazuyoshi; Kikuchi, Taiji; et al.
JAEA-Technology 2009-072, 144 Pages, 2010/03
"R&D Project on Irradiation Damage Management Technology for Structural Materials of Long-life Nuclear Plant" was carried out from FY2006 in a fund of a trust enterprise of the Ministry of Education, Culture, Sports, Science and Technology. The coupled irradiations or single irradiation by JOYO fast reactor and JRR-3 thermal reactor were performed for about two years. The irradiation specimens are very important materials to establish of "Evaluation of Irradiation Damage Indicator" in this research. For the acquisition of the examination specimens irradiated by the JOYO and JRR-3, we summarized about the overall plan, the work process and the results for the study to utilize these reactors and some facilities of hot laboratory (WASTEF, JMTR-HL, MMF and FMF) of the Oarai Research-and-Development Center and the Nuclear Science Research Institute in the Japan Atomic Energy Agency.
Hazama, Taira; Chiba, Go; Sato, Wakaei; Numata, Kazuyuki*
JAEA-Review 2009-003, 59 Pages, 2009/05
SLAROM-UF is a cell calculation code for fast reactors to produce effective cross sections with high accuracy in practical computing time, taking full advantage of fine and ultra-fine group calculation schemes. The fine group calculation scheme covers the whole energy range in a maximum of 900-group structure. Its group structure is finer above 52.5 keV with a minimum lethargy width of 0.008. Effective cross sections are evaluated based on the Bondarenko method. The ultra-fine group calculation scheme covers the energy range below about 52.5 keV. Its group structure is so fine ( 100,000 groups) as to treat resonance peaks as they are. Effective cross sections are calculated by solving an integral slowing down equation effectively, focusing only on elastic scattering and absorption reactions. Temperature can be specified freely by a user in the input data. The effective cross sections thus obtained are combined to calculate cell averaged cross sections.
Yokoyama, Kenji; Numata, Kazuyuki*; Hazama, Taira; Ishikawa, Makoto
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
The new solver for cross section adjustment and design accuracy evaluation has been developed for the new reactor physics analysis code system, MARBLE. In this development, object-oriented design was applied for achieving software extendibility. The new solver was successfully designed to easily add a uncertainty prediction method. This extendibility was confirmed by implementing the extended bias method. The new solver reproduces all functions of the conventional code system and can be used as standard solver for cross section adjustment and design accuracy evaluation in MARBLE.
Kugo, Teruhiko; Mori, Takamasa; Yokoyama, Kenji; Numata, Kazuyuki*; Ishikawa, Makoto
Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8 Pages, 2008/09
The extended bias factor methods, the LC and the PE methods, are applied to the prediction accuracy improvement for criticality and sodium void reactivity of a traditional two-region homogeneous fast reactor core with utilizing various experimental results. The extended bias factor methods are more effective than the conventional bias factor method. The PE method is more effective than the LC method when a small number of experimental results are used because of the advantage of the former method, its higher degree of freedom in combining experimental results. The advantage hardly affects on the prediction accuracy improvement when a sufficiently large number of experimental results are used. The variances due to cross sections originally included in the design calculation values for the criticality and sodium void reactivity are almost eliminated by the extended bias factor methods with use of about 200 experimental results regarding various neutronic characteristics. The uncertainty of the criticality is considerably reduced, because the uncertainty due to cross sections largely occupies in the original total uncertainty. The uncertainty reduction in the sodium void reactivity is not so much, because the uncertainty due to cross sections is smaller than that due to calculation methods.
Yokoyama, Kenji; Numata, Kazuyuki*
JAEA-Data/Code 2007-023, 39 Pages, 2008/01
A new cross section adjustment and nuclear design accuracy evaluation solver was developed as a set of modules for MARBLE (multi-purpose advanced reactor physics analysis system based on language of engineering). In order to enhance the system extendibility and flexibility, the object-oriented design and analysis technique was adopted to the development. In the new system, it is easy to add a new design accuracy evaluation method because a new numerical calculation module is independent from other modules. In order to validate the new solver, a test calculation was performed for a realistic calculation case of creating a new unified cross section library. In the test calculation, results calculated by the new solver agreed well with those by the conventional code system. Because the new solver implements all main functions of the conventional code system, MARBLE can be used as a new calculation code system for cross section adjustment and nuclear design accuracy evaluation.
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.
Chiba, Go; Numata, Kazuyuki*
Annals of Nuclear Energy, 34(6), p.443 - 448, 2007/06
In the present paper, we propose a neutron transport benchmark problem for fast critical assembly without homogenizations. With this problem, we can validate the applicability of neutron transport codes into highly-heterogeneous fast critical assembly analyses. In addition, this benchmark problem can be used to validate homogenization procedures for slab lattices.