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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.
Taniguchi, Yoshinori; Urano, Kenta; Mihara, Takeshi; Udagawa, Yutaka; Kakiuchi, Kazuo; Katsuyama, Jinya
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), p.1292 - 1301, 2025/10
Kakiuchi, Kazuo; Narukawa, Takafumi*; Udagawa, Yutaka; Katsuyama, Jinya; Mihara, Takeshi; Amaya, Masaki
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), p.1440 - 1449, 2025/10
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.
Katano, Ryota; Abe, Takumi; Cibert, H.*
JAEA-Research 2024-019, 22 Pages, 2025/05
JAEA-Research-2024-019.pdf:1.03MB
An accelerator-driven system (ADS) dedicated to transmutation of minor actinides (MAs) is driven in subcritical states. It is important for establishment of the subcriticality control of ADS to predict the burnup reactivity. To validate the prediction accuracy, the burnup reactivity, especially at the first cycle, must be measured with sufficient accuracy. In this study, we focus on Current-To-Flux (CTF) method. We have simulated the burnup reactivity monitoring during the ADS normal operation with the CTF method by performing fixed-source-burnup calculations using a continuous energy Monte Carlo code SERPENT2 with some tallies that models in-core fission chambers and have estimated its measurement uncertainty. We have clarified that the 10% biases of measure burnup reactivities appear independently of the burnup duration and their detector position dependence is particularly small in the outer region of the system.
Taniguchi, Yoshinori; Udagawa, Yutaka
IAEA-TECDOC-2053, p.94 - 96, 2024/05
Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 60(11), p.1386 - 1396, 2023/11
Times Cited Count：5 Percentile：61.58(Nuclear Science & Technology)The burnup calculations for estimating the nuclide composition of the spent fuel are highly dependent on nuclear data. Many nuclides in the latest version of the Japanese Evaluated Nuclear Data Library JENDL-5 were modified from JENDL-4.0 and the modification affects the burnup calculations. This study confirmed the validity of JENDL-5 in the burnup calculations. The PIE data of Takahama-3 was used for the validation. The effect of modifications of the parameters, e.g., cross sections and fission yields, from JENDL-4.0 to JENDL-5 on the nuclide compositions was quantitatively investigated. The calculation results showed that JENDL-5 has a similar performance to JENDL-4.0. The calculation results also revealed that the modifications of the cross sections of actinide nuclides, fission yields, and thermal scattering low data of hydrogen in HO affected the nuclide compositions of PWR spent fuels.
Narukawa, Takafumi; Hamaguchi, Shusuke*; Takata, Takashi*; Udagawa, Yutaka
Nuclear Engineering and Design, 411, p.112443_1 - 112443_12, 2023/09
Times Cited Count：1 Percentile：13.31(Nuclear Science & Technology)Narukawa, Takafumi; Hamaguchi, Shusuke*; Takata, Takashi*; Udagawa, Yutaka
Proceedings of Asian Symposium on Risk Assessment and Management 2022 (ASRAM 2022) (Internet), 11 Pages, 2022/12
Shaimerdenov, A.*; Gizatulin, S.*; Dyussambayev, D.*; Askerbekov, S.*; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sakaba, Nariaki
Nuclear Engineering and Technology, 54(8), p.2792 - 2800, 2022/08
Times Cited Count：14 Percentile：83.23(Nuclear Science & Technology)Riyana, E. S.; Okumura, Keisuke; Sakamoto, Masahiro; Matsumura, Taichi; Terashima, Kenichi
Journal of Nuclear Science and Technology, 59(4), p.424 - 430, 2022/04
Times Cited Count：1 Percentile：7.35(Nuclear Science & Technology)Narukawa, Takafumi
Nihon Genshiryoku Gakkai-Shi ATOMO
, 63(11), p.780 - 785, 2021/11
no abstracts in English
Narukawa, Takafumi; Udagawa, Yutaka
Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10
Ishitsuka, Etsuo; Mitsui, Wataru*; Yamamoto, Yudai*; Nakagawa, Kyoichi*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Nagasumi, Satoru; Takamatsu, Kuniyoshi; Kenzhina, I.*; et al.
JAEA-Technology 2021-016, 16 Pages, 2021/09
JAEA-Technology-2021-016.pdf:1.8MB
As a summer holiday practical training 2020, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the downsizing of reactor core were studied by the MVP-BURN. As a result, it is clear that a 1.6 m radius reactor core, containing 54 (18
3 layers) fuel blocks with 20% enrichment of U, and BeO neutron reflector, could operate continuously for 30 years with thermal power of 5 MW. Number of fuel blocks of this compact core is 36% of the HTTR core. As a next step, the further downsizing of core by changing materials of the fuel block will be studied.
Ikeda, Reiji*; Ho, H. Q.; Nagasumi, Satoru; Ishii, Toshiaki; Hamamoto, Shimpei; Nakano, Yumi*; Ishitsuka, Etsuo; Fujimoto, Nozomu*
JAEA-Technology 2021-015, 32 Pages, 2021/09
JAEA-Technology-2021-015.pdf:2.74MB
Burnup calculation of the HTTR considering temperature distribution and detailed burning regions was carried out using MVP-BURN code. The results show that the difference in k
, as well as the difference in average density of some main isotopes, is insignificant between the cases of uniform temperature and detailed temperature distribution. However, the difference in local density is noticeable, being 6% and 8% for U and Pu, respectively, and even 30% for the burnable poison 
B. Regarding the division of burning regions to more detail, the change of k is also small of 0.6%
k/k or less. The small burning region gives a detailed distribution of isotopes such as U, Pu, and B. As a result, the effect of graphite reflector and the burnup behavior could be evaluated more clearly compared with the previous study.
Ueta, Shohei; Sasaki, Koei; Arita, Yuji*
Nihon Genshiryoku Gakkai-Shi ATOMO, 63(8), p.615 - 620, 2021/08
no abstracts in English
Udagawa, Yutaka; Tasaki, Yudai
JAEA-Data/Code 2021-007, 56 Pages, 2021/07
JAEA-Data-Code-2021-007.pdf:5.05MB
Japan Atomic Energy Agency (JAEA) has released FEMAXI-8 in 2019 as the latest version of the fuel performance code FEMAXI, which has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly normal operation conditions and anticipated transient conditions. This report summarizes a newly developed model to analyze intragranular fission gas behaviors considering size distribution of gas bubbles and their dynamics. While the intragranular fission gas behavior models implemented in the previous FEMAXI versions have supported only treating single bubble size for a given fuel element, the new model supports multiple gas groups according to their size and treats their dynamic behaviors, making the code more versatile. The model was first implemented as a general module that takes arbitrary number of bubble groups and spatial mesh division for a given fuel grain system. An interface module to connect the model to FEMAXI-8 was then developed so that it works as a 2 bubble groups model, which is the minimum configuration of the multi-grouped model to be operated with FEMAXI-8 at the minimum calculation cost. FEMAXI-8 with the new intragranular model was subjected to a systematic validation calculation against 144 irradiation test cases and recommended values for model parameters were determined so that the code makes reasonable predictions in terms of fuel center temperature, fission gas release, etc. under steady-state and power ramp conditions.
Ho, H. Q.; Fujimoto, Nozomu*; Hamamoto, Shimpei; Nagasumi, Satoru; Goto, Minoru; Ishitsuka, Etsuo
Nuclear Engineering and Design, 377, p.111161_1 - 111161_9, 2021/06
Times Cited Count：3 Percentile：24.49(Nuclear Science & Technology)Nagase, Fumihisa; Narukawa, Takafumi; Amaya, Masaki
JAEA-Review 2020-076, 129 Pages, 2021/03
JAEA-Review-2020-076.pdf:3.9MB
Each light-water reactor (LWR) is equipped with the Emergency Core Cooling System (ECCS) to maintain the coolability of the reactor core and to suppress the release of radioactive fission products to the environment even in a loss-of-coolant accident (LOCA) caused by breaks in the reactor coolant pressure boundary. The acceptance criteria for ECCS have been established in order to evaluate the ECCS performance and confirm the sufficient safety margin in the evaluation. The limits defined in the criteria were determined in 1975 and reviewed based on state-of-the-art knowledge in 1981. Though the fuel burnup extension and necessary improvements of cladding materials and fuel design have been conducted, the criteria have not been reviewed since then. Meanwhile, much technical knowledge has been accumulated regarding the behavior of high-burnup fuel during LOCAs and the applicability of the criteria to the high-burnup fuel. This report provides a comprehensive review of the history and technical bases of the current criteria and summarizes state-of-the-art technical findings regarding the fuel behavior during LOCAs. The applicability of the current criteria to the high-burnup fuel is also discussed.
Yokoyama, Kenji; Lahaye, S.*
Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.109 - 116, 2020/10
CEA/DEN/DM2S/SERMA and JAEA/NSEC are working on benchmarks for burnup, isotopic concentrations and decay heat calculations in the collaboration framework between both organisms. Both actors of this benchmark are independently developing their own simulation code systems for computing quantities of interest in nuclear fuel cycle domain: MENDEL in CEA and MARBLE in JAEA. The purpose of the benchmark is to verify each system by comparing both calculation results on specific applications. MENDEL uses a several solvers for the resolution of Bateman equation. Runge-Kutta method or Chebyshev Rational Approximation method (CRAM) are used for irradiation computations. An analytical solver can also be used for decay calculations. MARBLE can use Krylov subspace method or CRAM method. As the first phase of the benchmark, we compared the calculated results of decay heat and isotropic concentrations following by a Pu-239 fast fission pulse. We applied nuclear data from three libraries: (1) JEFF-3.1.1, (2) JENDL/DDF-2015 + JENDL/FPY-2011, and (3) ENDF/B-VII.1. Nuclear data and burnup chain were generated from these libraries independently on each system. We confirmed that the results for both systems were in very good agreement with each other. Numerical results were also compared to experimental data. As the second phase of the benchmark, we are proceeding with a burnup calculation benchmark of MENDEL and MARBLE using the nuclear data and burnup chain provided by ORLIBJ33, which is a set of cross-section data based on JENDL-3.3 for ORIGEN-2 code system. We will also compare with calculation results by the ORIGEN-2 code with ORLIBJ33. Since the series of ORLIB, that is, ORLIBJ32, ORLIBJ33, and ORLIBJ40, have been widely used especially in Japan for many years, the comparison with ORLIB is effective for confirming the performance of MENDEL and MARBLE.
