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Maeda, Makoto; Komeda, Masao; Toh, Yosuke
Nuclear Science and Engineering, 200(1), p.181 - 194, 2026/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 199(6), p.1029 - 1043, 2025/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Ueki, Taro
Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025) (Internet), p.2018 - 2027, 2025/04
The criticality analysis of continuously mixed random media is essential to the safe retrieval of fuel debris. Image analysis of an oxide debris mockup reveals that the power spectrum cannot be fully explained by a single factor alone, but instead requires consideration of the complexity of multiple factors. This highlights the need for a randomized function capable of representing complex power spectra. To address this, we developed a new function called the Randomized Fourier Series (RFS), which introduces randomization in amplitude and phase. RFS allows the representation of power spectra with arbitrary shapes, facilitating realistic Monte Carlo (MC) simulations of random continuous material mixtures. For demonstration, taking the Lorentz power spectrum as an example, the spectrum flatness at low wavenumbers is analyzed to understand how the transition to white noise influences the fluctuation in neutron effective multiplication factor across independently generated random media replicas. Numerical results are presented for a mixture of 4 materials, along with the root mean-squared mass deviation over the constituent materials. The MC solver Solomon is employed with a partial volume pairing feature.
Watanabe, Tomoaki; Aizawa, Naoto*; Chiba, Go*; Tada, Kenichi; Yamamoto, Akio*
Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025) (Internet), p.288 - 297, 2025/04
Currently, a major burnup calculation method for the nuclide composition of nuclear fuel conducts neutron transport calculations at each burnup step to account for changes in the neutron spectrum. While this method is highly accurate, the large computational cost of neutron transport calculations can be problematic. Therefore, a fast burnup calculation method based on neutron spectrum reconstruction with the proper orthogonal decomposition (POD) and regression model is investigated. In this method, dimensionality reduction by POD is applied to many neutron fluxes obtained from detailed burnup calculations for various input parameter sets, and regression models are constructed to connect the dimensionality-reduced neutron fluxes and parameters. By substituting arbitrary input parameters to the regression models, the neutron flux is reconstructed and the burnup calculation is performed. This method performs burnup calculations that consider changes in the neutron spectrum based on input conditions without neutron transport calculations. The present method was applied to a PWR UO
fuel pin cell model. The results show the nuclide inventory can be calculated with a prediction accuracy within a few percent. In addition, it is found that the calculation error is dominated by the regression models, which implies the further improvement of the regression models leads to improving the accuracy.
Np, 
Am, and 
Am reaction rates in highly enriched uranium fuel cores at Kyoto University Critical AssemblyPyeon, C. H.*; Oizumi, Akito; Katano, Ryota; Fukushima, Masahiro
Nuclear Science and Engineering, 199(3), p.429 - 444, 2025/03
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Experimental analyses of neptunium-237 (Np), americium-241 (Am), and Am fission and Np capture reaction rates are conducted by the Serpent 2 code together with ENDF/B-VIII.0 and JENDL-5, using experimental data at neutron spectra of thermal and intermediate regions obtained in the solid-moderated and solid-reflected cores with highly-enriched uranium fuel at the Kyoto University Critical Assembly. Also, uncertainty quantification of fission and capture reaction rate ratios of test samples of Np, Am and Am with reference samples of uranium-235 (
U) and gold-197 (
Au) are evaluated by the MARBLE code system. In terms of fission reaction rate ratios of Np/U, Am/U and Am/U, a comparison between experiments and Serpent 2 calculations shows an accuracy about 5, 15 and 10%, respectively, together with ENDF/B-VIII.0 and JENDL-5. For capture reaction rate ratios of Np/Au, Serpent 2 calculations reveal a fairly good accuracy at the thermal neutron spectrum. The total uncertainties of Np/U, Am/U and Am/U fission reaction rate ratios by MARBLE with the covariance data of ENDF/B-VIII.0 and JENDL-5 are found to be about 4% at most in all cores, except for about 8% of Am/U with ENDF/B-VIII.0 at the intermediate neutron spectrum.
Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 199(1), p.18 - 41, 2025/01
Times Cited Count：2 Percentile：44.79(Nuclear Science & Technology)A series of integral experiments were conducted at FCA of JAEA, simulating LWR cores with a tight lattice cell of highly enriched MOX fuel containing more than 15% fissile plutonium. The three experimental configurations were constructed using foamed polystyrene with different void fractions to clarify the prediction accuracy of neutronic calculation codes and nuclear data among various neutron spectra. The nuclear characteristics measured in the experiments were criticality, moderator void reactivity worths, and sample reactivity worths. The preliminary analyses on experiments were conducted using a deterministic calculation code conventionally used for fast reactors with JENDL-4.0. Most reactivity worth calculations correlated well with the experimental values. Specifically for the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations were performed with MVP3 code by modeling the experimental setup in detail, confirming the validity of the deterministic calculations.
Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 17 Pages, 2025/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)
fuel in an intermediate spectrum sodium-cooled reactor with a lower linear heat ratingKuwagaki, Kazuki; Chikazawa, Yoshitaka; Yan, X.
Nuclear Science and Engineering, 12 Pages, 2025/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Katano, Ryota; Fukushima, Masahiro; Pyeon, C. H.*
Nuclear Science and Engineering, 11 Pages, 2025/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Proposed here is a novel method involving integral experiments effective in reducing nuclear-data-induced uncertainty. The method is formulated by the Extended Bias Factor method and the L1-norm-based sparse modeling to address computational difficulty in combinational optimization. The pseudo-design parameters only sensitive to specific microscopic reactions are defined for the validation of nuclear data. The method is applied to two pseudo parameters: neptunium-237 capture and bismuth-209 inelastic scattering cross sections, while considering integral experimental data used in ADJ2017 together with sample worth measurements made at the Kyoto University Critical Assembly. The results indicate that the proposed method successfully identifies a small subset of effective integral experiments.
Watanabe, Tomoaki; Suyama, Kenya; Tada, Kenichi; Ferrer, R. M.*; Hykes, J.*; Wemple, C. A.*
Nuclear Science and Engineering, 198(11), p.2230 - 2239, 2024/11
Times Cited Count：1 Percentile：23.55(Nuclear Science & Technology)A new nuclear data library for the advanced lattice physics code CASMO5 has been prepared based on JENDL-5. In JENDL-5, many essential nuclides for conventional LWR analysis have also been modified based on state-of-the-art evaluations. The new JENDL-5-based CASMO5 library was prepared by replacing as much of the nuclear data of the current CASMO5 ENDF/B-VII.1-based library as possible with JENDL-5. This study verified and validated the new library. Verifications were performed based on the OECD/NEA burnup credit criticality safety benchmark phase III-C, and the calculated k
and fuel compositions of the BWR fuel assembly were compared with reported benchmark results. Comparison with the MCNP6.2 result was also performed using the same benchmark model. In addition, the TCA critical experiment and Takahama-3 post-irradiation experiment were used for validation. The results indicate that the new library performs well and is comparable to the ENDF/B-VII.1-based library in predictions of reactivity and fuel compositions for LWR systems.
Katano, Ryota; Oizumi, Akito; Fukushima, Masahiro; Pyeon, C. H.*; Yamamoto, Akio*; Endo, Tomohiro*
Nuclear Science and Engineering, 198(6), p.1215 - 1234, 2024/06
Times Cited Count：1 Percentile：13.88(Nuclear Science & Technology)In this study, we have demonstrated that data assimilation using lead and bismuth sample reactivities measured in the Kyoto University Critical Assembly A-core can successfully reduce the uncertainty of the coolant void reactivity in accelerator-driven systems derived from inelastic-scattering cross-sections of lead and bismuth. We re-evaluated and highlighted the experimental uncertainties and correlations of the sample reactivities for the data assimilation formula. We used the MCNP6.2 code to evaluate the sample reactivities and their uncertainties, and performed data assimilation using the reactor analysis code system MARBLE. The high-sensitivity coefficients of the sample reactivities to lead and bismuth allowed us to reduce the cross-section-induced uncertainty of the void reactivity of the accelerator-driven system from 6.3% to 4.8%, achieving a provisional target accuracy of 5% in this study. Furthermore, we demonstrated that the uncertainties arising from other dominant factors, such as minor actinides and steel, can be effectively reduced by using integral experimental data sets for the unified cross-section dataset ADJ2017.
Tuya, D.; Nagaya, Yasunobu
Nuclear Science and Engineering, 198(5), p.1021 - 1035, 2024/05
Times Cited Count：1 Percentile：13.88(Nuclear Science & Technology)In Monte Carlo neutron transport calculations for local response or deep penetration problems, some estimation of an importance function is generally required in order to improve their efficiency. In this work, a new recursive Monte Carlo (RMC) method, which is partly based on the original RMC method, for estimating an importance function for local variance reduction (i.e., source-detector type) problems has been developed. The new RMC method has been applied to two sample problems of varying degrees of neutron penetrations, namely a one-dimensional iron slab problem and a three-dimensional concrete-air problem. The biased Monte Carlo calculations with variance reduction parameters based on the obtained importance functions by the new RMC method have been performed to estimate detector responses in these problems. The obtained results are in agreement with those by the reference unbiased Monte Carlo calculations. Furthermore, the biased calculations offered an increase in efficiency on the order of 1 to 10
in terms of the figure of merit (FOM). The results also indicated that the efficiency increased as the neutron penetration became deeper.
TaEndo, Shunsuke; Kimura, Atsushi; Nakamura, Shoji; Iwamoto, Osamu; Iwamoto, Nobuyuki; Rovira Leveroni, G.; Toh, Yosuke; Segawa, Mariko; Maeda, Makoto
Nuclear Science and Engineering, 198(4), p.786 - 803, 2024/04
Times Cited Count：1 Percentile：13.88(Nuclear Science & Technology)Pyeon, C. H.*; Katano, Ryota; Oizumi, Akito; Fukushima, Masahiro
Nuclear Science and Engineering, 197(11), p.2902 - 2919, 2023/11
Times Cited Count：3 Percentile：42.83(Nuclear Science & Technology)Sample reactivity and void reactivity experiments are carried out in the solid-moderated and solid-reflected cores at the Kyoto University Critical Assembly (KUCA) with the combined use of aluminum (Al), lead (Pb) and bismuth (Bi) samples, and Al spacers simulating the void. MCNP6.2 eigenvalue calculations together with JENDL-4.0 provide good accuracy of sample reactivity with the comparison of experimental results; also experimental void reactivity is attained by using MCNP6.2 together with JENDL-4.0 and ENDF/B-VII.1 with a marked accuracy of relative difference between experiments and calculations. Uncertainty quantification of sample reactivity and void reactivity is acquired by using the sensitivity coefficients based on MCNP6.2/ksen and covariance library data of SCALE6.2 together with ENDF/B-VII.1, arising from the impact of uncertainty induced by Al, Pb and Bi cross sections. A series of reactivity analyses with the Al spacer simulating the void demonstrates the means of analyzing the void in the solid-moderated and solid-reflected cores at KUCA
Bess, J. D.*; Chipman, A. S.*; Pope, C. L.*; Jensen, C. B.*; Ozawa, Takayuki; Hirooka, Shun; Kato, Masato*
Nuclear Science and Engineering, 197(8), p.1845 - 1872, 2023/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Pretransient characterization was performed for the EBR-II MOX fuel pellets from the SPA-2/-2B Operational Reliability Testing collaboration between Japan and US. The continued collaboration will investigate the transient performance of these rods in TREAT at Idaho National Laboratory. The results will fill a gap in existing transient performance data for MOX as these rods have a peak burnup of ~134.4 GWd/t in the EBR-II. Fuel pellet properties were gathered from available resources and their irradiation and decay history evaluated. Further reactor physics calculations were performed to support the experiment design, reactor operations, and safety analyses necessary to enable the programmatic success. Of the three irradiated fuel pins, two will undergo transient testing, and all three will undergo post-irradiation examination.
Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*
Nuclear Science and Engineering, 197(8), p.2251 - 2257, 2023/08
Times Cited Count：2 Percentile：21.09(Nuclear Science & Technology)
variations by nonuniform core configurationsGunji, Satoshi; Araki, Shohei; Suyama, Kenya
Nuclear Science and Engineering, 197(8), p.2017 - 2029, 2023/08
Times Cited Count：1 Percentile：13.88(Nuclear Science & Technology)The fuel debris generated by the accident at the Tokyo Electric Power Company's Fukushima Daiichi Nuclear Power Plant is expected to have not only heterogeneous but also nonuniform compositions. Similarly, damaged fuel assemblies remaining in the reactor vessels also have nonuniform configurations due to some missing fuel rods. This non-uniformity may cause changing neutron multiplication factors. The effect of non-uniformity on the neutron multiplication factor is clarified by computations, and the possibility of experimentally validating the computations used for criticality management is being investigated. For this purpose, in this study the criticality effects of several core configurations of a new critical assembly, STACY, of the Japan Atomic Energy Agency with nonuniform arrangements of uranium oxide fuel rods, concrete rods, and stainless-steel rods were studied to confirm benchmarking potential. The difference in these arrangements changed the neutron multiplication factor by more than 1 $. We confirmed that changes in local neutron moderation conditions and the clustering of specific components caused this effect. In addition, the feasibility of benchmark experimental cores with nonuniform arrangements is evaluated. If benchmarking of such experiments could be realized, it would help to validate calculation codes and to develop criticality management methods by machine learning.
Sugawara, Takanori; Kunieda, Satoshi
Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2023) (Internet), 7 Pages, 2023/08
This study investigates the impact of the change from JENDL-4 to JENDL-5 on neutronics analysis of transmutation systems. As the transmutation systems, the following two systems are targeted: JAEA-ADS, a lead-bismuth cooled accelerator-driven system, and MARDS, a molten salt chloride accelerator-driven system. For the JAEA-ADS, the k-eff value increased 189 pcm from JENDL-4 to JENDL-5. It was found that the revisions of various nuclides affected to this difference. For example, the revision of 
N indicated an increase of 200 pcm from the JENDL-4 result. For the MARDS, it was found that the major revision of 
Cl and 
Cl cross sections was the main cause of the k-eff differences. This study confirmed that the difference in the nuclear data libraries still indicated differences in calculation results for the transmutation systems.
Kondo, Ryoichi; Nagaya, Yasunobu
Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2023) (Internet), 10 Pages, 2023/08
A functional expansion tally (FET) method with numerical basis functions generated by singular value decomposition (SVD) is newly proposed. Traditionally, analytical functions were used for the FET calculations, e.g., Legendre polynomials for a one-dimensional distribution. However, the expansion terms could increase to reconstruct steep or complex distributions with these functions. A basis set that can well represent the target distribution with lower order expansion is desired to achieve high accuracy with the small computational resource. In the present study, a numerical basis set is generated from snapshot data by using SVD. This approach is based on the reduced order modeling (ROM). We applied ROM to the FET method in Monte Carlo calculations. The numerical result showed the applicability of the proposed method, on the other hand, some issues were revealed, e.g., discretization of the snapshot data.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki
Journal of Nuclear Engineering and Radiation Science, 9(3), p.031401_1 - 031401_11, 2023/07
In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) investigated in JAEA, the use of a specific fuel assembly with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR. Since the fuel rods have an asymmetric layout by the inner duct, the validity confirmation of the numerical results of an in-house subchannel analysis code named ASFRE was required. In this paper, therefore, the code-to-code comparisons was applied with numerical results of ASFRE and those of an in-house CFD code named SPIRAL. The applicability of ASFRE was indicated through the confirmation of the consistency of specific temperature distributions.
