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Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 63(1), p.3 - 18, 2026/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study investigated the impact of nuclear data updates from JENDL-4.0 (J4) to JENDL-5 (J5) on the light-water reactor fuel burnup calculations. Burnup calculations were conducted with J4 and J5 for PWR pin-cell and BWR fuel assembly geometries. The calculation results revealed significant burnup-dependent differences in the neutron multiplication factor (k
). Across the burnup range of 0-50 GWd/t, k
values of J5 were consistently smaller than those of J4 and the difference gradually increased as burnup progressed. Direct sensitivity calculations, in which each nuclide data was replaced from J4 to J5, indicated that updates to the cross-sections of 
U, 
U, and 
Pu and the thermal scattering law data of H in H
O notably impacted the k differences. For the BWR assembly geometry containing Gd fuels, large k differences were observed in the burnup range of 10-15 GWd/t. This difference was primarily attributed to updates in the U, 
Gd, and 
Gd cross-sections, and thermal scattering law data of H in HO. Furthermore, we investigated how the nuclear data updates affected the k differences by examining nuclide number densities, the energy-dependent sensitivities, and the neutron spectra.
Nguyen, H. H.
Annals of Nuclear Energy, 218, p.111361_1 - 111361_9, 2025/08
Times Cited Count：1 Percentile：50.28(Nuclear Science & Technology)This study examined the criticality characteristics of a partially damaged reactor model, in which fuels located at the core center melt into fuel debris of varying shapes, while fuels situated at the core edges remain intact. The investigation was conducted using the Serpent code with the JENDL-5 nuclear data library. The results of the calculations indicate that when the volume of fuel debris is small and maintained at a constant level, the shape of the fuel debris does not result in significant alterations in the variation law of k
of the system. In contrast, for the scenario in which the volume of the fuel debris is variable, the k variation law can be divided into two groups for the reference case with a system temperature of 300 K and no boron in the water. The first group comprises fuel debris with shapes that are cuboid and cylindrical, while the second group comprises fuel debris with shapes that are spherical, cone-shaped, and truncated cone-shaped.
and -MOX lattice calculationsFujita, Tatsuya
Journal of Nuclear Science and Technology, 62(8), p.731 - 739, 2025/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study estimated the influence of implicit effect on the k-infinity uncertainty in the PWR-UO and -MOX fuel lattice geometries. Firstly, the preliminary investigation was performed, where the influence of implicit effect was roughly estimated based on the sandwich formula using the cross-section (XS) covariance matrix and the sensitivity coefficient. It was confirmed that the influence of implicit effect became large in the fission and (n,
) reactions of heavy nuclides and the change of this dependence was small for the burnup of UO and MOX fuel assemblies. Then, focussing on the heavy nuclides, the influence of implicit effect was compared under several energy group conditions of the XS covariance matrix and neutron transport calculation. For Pu and 
Pu, the noticeable influence of implicit effect was observed in MOX fuel pin-cell geometry. However, increasing the number of energy groups for neutron transport calculations and that of the XS covariance matrix can reduce the influence of implicit effect. Consequently, by appropriately setting the number of energy groups for neutron transport calculations and that of the XS covariance matrix, it became practically possible not to explicitly consider the implicit effect during the random sampling.
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.
Fujita, Tatsuya
Journal of Nuclear Science and Technology, 62(5), p.470 - 479, 2025/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study confirmed the efficiency of a combined approach of the control variates (CV) and the Latin hypercube sampling (LHS), which enhanced the random-sampling-based uncertainty quantification due to cross-section (XS) covariance data, by considering the effect of statistical variation and also performed the sensitivity analyses on the influence due to the selection of alternative parameter to apply CV. The convergence performance for the uncertainty of infinite multiplication factor (k-infinity) during the random sampling was compared between several efficient sampling techniques such as the antithetic sampling (AS), LHS, CV, and the combined approaches of them in the PWR-UO fuel assembly geometry. The k-infinity uncertainty was evaluated by statistically processing several times Serpent2 calculations using perturbed ACE-formatted XS files based on ENDF/B-VIII.0. CV+LHS was more efficient than AS, LHS, and CV+AS. In addition, sensitivity analyses were performed to select alternative parameters used in CV. The 3
3 mini fuel lattice calculation can improve the efficiency of CV+LHS. The reason was qualitatively considered that this calculation can capture the influence of XS covariance data for Gd isotopes. Consequently, the applicability of CV+LHS for the improvement of convergence performance to evaluate the k-infinity uncertainty during the random sampling was confirmed.
Fukuda, Kodai
Annals of Nuclear Energy, 208, p.110748_1 - 110748_10, 2024/12
Times Cited Count：1 Percentile：20.46(Nuclear Science & Technology)Fujita, Tatsuya
Proceedings of Best Estimate Plus Uncertainty International Conference (BEPU 2024) (Internet), 14 Pages, 2024/05
The uncertainty analysis of PWR depletion test problem on the OECD/NEA/NSC LWR-UAM benchmark Phase II based on JENDL-5 was performed as a preliminary investigation. The random sampling was used to quantify the uncertainty of k-infinity and nuclide inventories, the cross section (XS), the fission product yield (FPY), the decay constant, and the decay branch ratio were randomly perturbed, and several times of SERPENT 2.2.1 calculations were performed. XSs in the ACE file were perturbed by the ACE file perturbation tool using FRENDY with the 56-group covariance matrix generated by NJOY2016.72. The perturbation quantity of independent FPY was evaluated using the FPY covariance matrix prepared in JENDL-5, and the perturbed cumulative FPY was reconstructed based on the relationship between the independent and cumulative FPYs. The decay constant was independently perturbed for each nuclide. To perturb the decay branch ratios, the covariance matrix was generated by applying the generalized least square method and randomly perturbed based on this covariance matrix in the same manner as the independent FPY. In general, the influence due to decay data was an order of magnitude smaller than the influences due to XS and FPY uncertainties. For the uncertainty of k-infinity and transuranic nuclide inventories, the influence due to XS uncertainty was dominant, and that due to FPY and decay data uncertainties was one or a few orders of magnitude smaller. On the other hand, for the uncertainty of FP nuclide inventories, the influence due to FPY uncertainty was almost the same or larger than that due to XS uncertainty. It was also confirmed that the influence due to either XS or FPY uncertainty became different for each FP nuclide. In future studies, the influence due to XS uncertainty on FP nuclides will be discussed because it was not prepared in JENDL-5 and not considered in the present paper.
Fujita, Tatsuya
Proceedings of International Conference on Physics of Reactors (PHYSOR 2024) (Internet), p.718 - 727, 2024/04
The convergence process of the k-infinity uncertainty during random-sampling-based uncertainty quantification was compared between several efficient sampling techniques. The k-infinity uncertainty was evaluated by statistically processing several times of SERPENT 2.2.1 calculations using perturbed ACE files based on JENDL-5 cross-section covariance data. The antithetic sampling (AS), the Latin hypercube sampling (LHS), the control variates (CV), and the combination approaches of them were focused on in the present paper. In PWR-UO fuel assembly geometry without the nuclide depletion, as discussed in past studies, AS and LHS showed higher efficient convergence than nominal sampling without any efficient sampling techniques. In terms of CV, though a stand-alone application did not have a large impact on the k-infinity uncertainty convergence, its performance was improved in combination with AS, as discussed in the past study. In addition, a new combined approach of LHS and CV (CV+LHS) was proposed in the present paper. CV+LHS improved the k-infinity uncertainty convergence and was more efficient than CV+AS. The main reason for this improvement was that the convergence for the mean value of alternative parameters in CV was enhanced by applying LHS. Consequently, this study proposed the new combined approach of CV+LHS and confirmed its efficiency performance for the random-sampling-based uncertainty quantification in the PWR-UO fuel assembly geometry. The applicability of CV+LHS for the nuclide-depletion calculations will be confirmed in future studies.
Kato, Aitaro*; Saiga, Atsushi; Takeda, Tetsuya*; Iwasaki, Takaya*; Matsuzawa, Toru*
Earth, Planets and Space (Internet), 66(1), p.86_1 - 86_8, 2014/12
Times Cited Count：26 Percentile：57.99(Geosciences, Multidisciplinary)To understand the mechanism of an intensive non-volcanic seismic swarm in the Kii Peninsula, Japan, we used a dense seismic linear array to measure fine-scale variations of seismic velocities and converted teleseismic waves. A low-velocity anomaly confined to just beneath the seismic swarm area is clearly imaged, which spatially correlates with an uplifted surface area, and a highly conductive and strong attenuative body. These results suggest that fluids such as partial melt or water are present beneath this non-volcanic seismic swarm area. It is notable that the island arc Moho below the seismic swarm area is at depths of ca. 32 km in the northern part of the seismic swarm area, and shallows to ca. 20 km towards the south, due to an upwardly raised structure of serpentinized mantle wedge. In addition, we show that hydrated oceanic crust of the subducting Philippine Sea slab is characterized by low-velocities with a high Poisson's ratio at depths shallower than 40 km. Water released from the subducting oceanic crust could cause serpentinization of the mantle wedge and infiltration into the forearc base of the overlying plate. The interaction between dehydration of the subducting oceanic crust and hydration of the mantle wedge and overlying plate exerts an important role in driving the non-volcanic seismic swarm activity in the Kii Peninsula.
Dwijayanto, A. P.*; Nishihara, Kenji; Okamura, Tomohiro*; Nakase, Masahiko*
no journal, ,
NMB4 is a nuclear fuel cycle simulator capable of modelling various nuclear fuel cycle scenarios from front-end to back-end. However, its capability to simulate a Molten Chloride Fast Reactor (MCFR), a subset of Molten Salt Reactor (MSR), had yet to be tested. Verifying NMB4 capability to model MCFR fuel cycle in a nuclear energy system is crucial for determining the limit of applicability and future development of the NMB4 simulator. A mass balance comparison between the NMB4 nuclear fuel cycle simulator and the Monte Carlo code Serpent-2 was then performed. A long-lived marine-based MCFR core with minimal reprocessing was modeled in a Monte-Carlo code as the closest scheme with the depletion capability supported by NMB4. From the results, NMB4 was found to generate an acceptable mass balance compared to Serpent-2, and a simple case study using a marine-based MCFR was subsequently simulated to analyse the mass balance and the disposal area required in a once-through cycle. Several discrepancies in code verification have been discussed, and future developments are suggested.
Fujita, Tatsuya
no journal, ,
Past studies have mentioned that the treatment of implicit effect for cross-section perturbation affects the sensitivity coefficients and then the uncertainty analysis results for the k-infinity. The several approaches to consider the above implicit effect has also been discussed for the random-sampling-based uncertainty analysis. In this study, the influence due to implicit effect on typical nuclides and nuclear reactions in PWR 1717 UO and MOX fuel assemblies was confirmed prior to treat the implicit effect on the random-sampling-based uncertainty analysis in future studies. In the UO fuel assembly, the influence due to implicit effect on the k-infinity was small, thus the uncertainty quantification only considering the explicit effect would be applicable. On the other hand, further discussion about the influence due to implicit effect was necessary for the MOX fuel assembly.
Fujita, Tatsuya
no journal, ,
Past studies have mentioned that the treatment of implicit effect for cross-section perturbation affects the sensitivity coefficients and then the uncertainty propagation for neutronics parameters. The several approaches to consider the implicit effect have also been discussed for the random-sampling-based uncertainty propagation. In this study, continued from the presentation in the AESJ 2024 spring meeting, the influence of implicit effect for multi-group cross-section perturbation on random-sampling-based uncertainty propagation was evaluated for the k-infinity uncertainty in the multi-group neutron transport calculation. While the influence of the implicit effect was small on the UO fuel pin-cell geometry, large influence was observed in the MOX fuel pin-cell geometry. However, it was considered that this influence could be reduced by refining the number of energy groups in neutron transport calculation and cross-section covariance matrix.
Yoshikawa, Tomoki; Izawa, Kazuhiko; Araki, Shohei; Watanabe, Tomoaki; Gunji, Satoshi
no journal, ,
We conducted performance tests on five types of cores at the modified STACY as first critical cores for resumption of operation. The performance test consisted of a core with 253 and 277 fuel rods and 253 fuel rods loaded with a sample drive at grid intervals of 1.5 cm, and 213 and 241 fuel rods at grid intervals of 2.54 cm. We analyzed the critical water level in each core using MVP with JENDL-4 as the nuclear data for the configuration. Similar calculations were conducted for the other codes MCNP, Solomon and SERPENT to compare the evaluation results.
Szogradi, M.*; Nagaya, Yasunobu
no journal, ,
A new nodal diffusion solver Ants has been developed at VTT for the development of Generation IV reactor concepts. The two-step approach where group constants are generated via Monte Carlo (MC) methods is used for the reactor core analysis. To validate this approach, the JOYO MK-I core published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments was analyzed. The Serpent Monte Carlo code was used to obtain group constants of the JOYO MK-I core. The Serpent few-group cross sections were then used as input data for Ants. Calculations were also carried out with Serpent and the MVP Monte Carlo code for a full-core 3D model to assess the general performance of the code comparing neutron spectrums and power spatial distributions. Results obtained with the 3D Ants JOYO model were compared with the MC results as the first assessment of the nodal diffusion solver with fast system.
