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Kwon, Saerom*; Konno, Chikara; Honda, Shogo*; Kenjo, Shunsuke*; Sato, Satoshi*
Fusion Engineering and Design, 223, p.115548_1 - 115548_8, 2026/02
In order to evaluate the accuracy of the iron data in the latest nuclear data libraries (FENDL-3.2b, JENDL-5, ENDF/B-VIII.0 and JEFF-3.3) used in the fusion neutron source design, we performed their benchmark tests by using QST/TIARA iron experiment with quasi mono-energy neutrons of 40 and 65 MeV and JAEA/FNS iron experiment with DT neutrons. From the test results, we have found the following issues; (1) The calculation results with FENDL-3.2b underestimate the measured neutron fluxes of the continuous energy range (10-60 MeV) by a factor of 0.6 in the TIARA experiment with 65 MeV neutrons; (2) The calculation results with FENDL-3.2b tend to underestimate the measured neutron flux above 10 MeV by a factor of 0.8 at depth of 70 cm and overestimate the measured ones below 10 keV by a factor of 1.3 up to depth of 40 cm in the FNS experiment. We investigated those issues in detail and clarified their reasons.
Yanagisawa, Hiroshi; Motome, Yuiko
JAEA-Research 2025-010, 197 Pages, 2025/11
JAEA-Research-2025-010.pdf:3.5MB
For understandings of nuclear criticality risks of TRIGA fuel rods and review of safety measures for handling them, nuclear criticality characteristics for infinite and finite heterogeneous lattice systems composed of the NSRR fuel rods were re-evaluated with the use of a detailed computational model for the fuel rod. The MVP version 3 code was used with the JENDL libraries including the latest version, JENDL-5, for the re-evaluation. As the criticality characteristics, variations of neutron multiplication factors of the infinite and water-reflected finite systems were examined in detail with parameters of the lattice pitch and density of moderator water. From the results of the re-evaluated criticality characteristics, the minimum critical number of fuel rods for the water-reflected hexagonal shaped lattice system was obtained to be 46.8 
0.2 using the JENDL-5 library. Moreover, the attainability of criticality without the water as moderator and reflector was examined because the zirconium hydride moderator and graphite reflector are equipped with the TRIGA fuel rod. It was found that the criticality is possible to be attained by 115.7 0.6 of the number of fuel rods, which is the smaller number of fuel rods than loaded in the NSRR standard core, even though no water exists.
Yanagisawa, Hiroshi; Motome, Yuiko
JAEA-Research 2025-001, 99 Pages, 2025/06
JAEA-Research-2025-001.pdf:1.98MB
The detailed computational models for nuclear criticality analyses on the first startup cores of NSRR (Nuclear Safety Research Reactor), which is categorized as a TRIGA-ACPR (Annular Core Pulse Reactor), were created for the purposes of deeper understandings of safety inspection data on the neutron absorber rod worths of reactivity and improvement of determination technique of the reactivity worths. The uncertainties in effective neutron multiplication factor (k
) propagated from errors in the geometry, material, and operation data for the present models were evaluated in detail by using the MVP version 3 code with the latest Japanese nuclear data library, JENDL-5, and the previous versions of JENDL libraries. As a result, the overall uncertainties in k for the present models were evaluated to be in the range of 0.0027 to 0.0029 
k. It is expected that the present models will be utilized as the benchmark on k for TRIGA-ACPR. Moreover, it is confirmed that the overall uncertainties were sufficiently smaller than the values of absorber rod worths determined in NSRR. Thus, it is also considered that the present models are applicable to further analyses on the absorber rod worths in NSRR.
Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 16 Pages, 2025/04
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.
and -MOX lattice calculationsFujita, Tatsuya
Journal of Nuclear Science and Technology, 9 Pages, 2025/03
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.
Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 17 Pages, 2025/00
Konno, Chikara
JAEA-Conf 2024-002, p.80 - 85, 2024/11
The official ACE files of JENDL-5 were released in December, 2022. The neutron ACE file of JENDL-5 was mainly produced with the FRENDY code, while the data on nuclear heating and damage (heating number, damage production energy) were done with the NJOY2016.65 code modified for JENDL-5. This presentation explains the modified points of NJOY2016.65 and the data on nuclear heating and damage in the neutron ACE file of JENDL-5.
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：30.56(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.
Okita, Shoichiro; Aoki, Takeshi; Fukaya, Yuji; Tachibana, Yukio
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 5 Pages, 2024/11
Okita, Shoichiro; Abe, Yutaka*; Tasaki, Seiji*; Fukaya, Yuji
Radioisotopes, 73(3), p.233 - 240, 2024/11
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.
Fukushima, Masahiro; Okajima, Shigeaki*; Mukaiyama, Takehiko*
Journal of Nuclear Science and Technology, 61(4), p.478 - 497, 2024/04
Times Cited Count：3 Percentile：59.67(Nuclear Science & Technology)A series of integral experiments was conducted to evaluate the fission and the capture cross- sections of transuranic (TRU) nuclides at the fast critical facility FCA of the Japan Atomic Energy Agency (JAEA). The experiments were carried out using seven uranium-fueled assemblies of the FCA. The neutron energy spectra of the core regions were adjusted so as to change from an intermediate neutron spectrum to a fast neutron spectrum on an assembly-by-assembly basis. The integral data measured with these experimental configurations provide some neutron energy characteristics: 1) fission rate ratios (FRRs) of 
Np, Pu, 
Pu, 
Am, 
Am, and 
Cm relative to Pu by using absolutely calibrated fission chambers, 2) small sample reactivity worths (SRWs) of Np, Pu, Pu, Am, and Am where oxide powders of around 15 to 20 grams were used, 3) criticalities, and 4) spectral indices such as fission rate ratios of U relative to U. In this paper, details of the SRW measurements are reported, and the latest Japanese Evaluated Nuclear Data Library JENDL-5 is tested by using the integral data obtained in systematically varied neutron energy spectra.
Konno, Chikara; Kochiyama, Mami; Hayashi, Hirokazu
Mechanical Engineering Journal (Internet), 11(2), p.23-00386_1 - 23-00386_11, 2024/04
Activation cross-section libraries for the ORIGEN and ORIGEN-S codes have been generated from JENDL-5 and JENDL/AD-2017. The ORIGEN activation cross-section libraries of the 200 and 48 group structures were generated with the AMPX-6 code, while the ORIGEN-S activation cross-section libraries with a MAXS format of the 199 group structure were done with the PREPO2018 code. Activation calculations for JPDR were carried out in order to validate the produced ORIGEN and ORIGEN-S activation cross-section libraries. The following comparisons were performed: the ORIGEN calculation results with the produced activation cross-section libraries and bundled ones, the 200 group and 48 group ORIGEN calculations, the ORIGEN and ORIGEN-S calculations with the JENDL-5 activation cross-section libraries, etc. Most of the differences of the calculation results were less than 20%, which demonstrated that the libraries were produced adequately.
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.
Konno, Chikara
JAEA-Conf 2023-001, p.143 - 146, 2024/02
I modified NJOY2016.67 to produce photonuclear ACE files which can be used in MCNP6.2 and PHITS3.27 and produced the ACE file of the JENDL-5 photonuclear sub-library. Simple test calculations with the produced ACE file supported that the produced ACE file had no serious problems.
Suzuki, Motomu*; Nagaya, Yasunobu
Journal of Nuclear Science and Technology, 61(2), p.177 - 191, 2024/02
Times Cited Count：2 Percentile：33.46(Nuclear Science & Technology)With the release of the latest Japanese evaluated nuclear data library JENDL-5, the prediction accuracy of JENDL-5 for neutronics parameters of the BEAVRS benchmark for the hot zero power condition was evaluated in this study. The criticality, control rod bank worth (CRW), isothermal temperature coefficient (ITC), and in-core detector signals were calculated and compared with the measured data for evaluation. For the criticality, the calculation-to-experimental (C/E) values varied between 1.0001 and 1.0045. Sensitivity analysis by replacing cross section data from the JENDL-4.0u1 with JENDL-5 revealed that 
H, U, U, and 
O significantly affected the criticality. The individual CRW agreed within 50 pcm, and total CRW also agreed within 100 pcm from the measured values. The ITC results calculated with a temperature deviation of 5.56 K case were negatively overestimated comparing with the measured values; whereas those of with 2.78 K were improved and agree with the measured values within a standard deviation. The axial detector signals indicated a maximum relative error of 4.46% and the root mean squared error (RMSE) of 2.13%. The differences between the previous version of JENDL-4.0u1 and JENDL-5 were also investigated.
Tada, Kenichi; Nagaya, Yasunobu; Taninaka, Hiroshi; Yokoyama, Kenji; Okita, Shoichiro; Oizumi, Akito; Fukushima, Masahiro; Nakayama, Shinsuke
Journal of Nuclear Science and Technology, 61(1), p.2 - 22, 2024/01
Times Cited Count：13 Percentile：95.11(Nuclear Science & Technology)The new version of the Japanese evaluated nuclear data library, JENDL-5, was released in December 2021. This paper demonstrates the validation of JENDL-5 for fission reactor applications. Benchmark calculations are performed with the continuous-energy Monte Carlo codes MVP and MCNP and the deterministic code system MARBLE. The benchmark calculation results indicate that the performance of JENDL-5 for fission reactor applications is better than that of the former library JENDL-4.0.
Nakayama, Shinsuke
Journal of Nuclear Science and Technology, 60(12), p.1447 - 1453, 2023/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The d+Be neutron source is a candidate for transportable neutron source for on-site nondestructive inspection of infrastructure facilities such as bridges, tunnels and so on. The applicability of the d+Be neutron source to a transportable fast neutron source is explored by Monte Carlo particle transport simulations with PHITS and JENDL-5. The simulation results show that by increasing the shielding thickness by about 1.5 times, it is possible to realize the d+Be neutron source with the comparable performance to another candidate, the 2.5-MeV p+Li neutron source, at lower beam energy.
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：68.50(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.
Kawaguchi, Maho*; Shiba, Shigeki*; Iwahashi, Daiki*; Okawa, Tsuyoshi*; Gunji, Satoshi; Izawa, Kazuhiko; Suyama, Kenya
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 8 Pages, 2023/10
The Nuclear Regulation Authority (NRA) has been working on an experimental approach for evaluating the criticality of fuel debris produced by the Fukushima Daiichi Nuclear Power Plant (FDNP) accident since 2014, collaborating with the Japan Atomic Energy Agency (JAEA). As part of the approach, JAEA has modified the STAtic experiment Critical facilitY (STACY) for critical experiments to evaluate characteriscs of pseudo-fuel debris. As the preliminary analyses, we verified critical characteristics with major nuclear data libraries for the proposed core configuration patterns. The three-dimensional continuous-energy Monte Carlo neutron and photon transport code, SERPENT-V2.2.0 was used with the latest JENDL, JENDL-5. As a result, larger multiplication factors of JENDL-5 across the modified STACY core configuration patterns were evaluated in comparison to the other libraries. And, 
H scattering and U fission sensitivity coefficients of JENDL-5 were different from those of the other libraries. Comparing among analyses with those libraries, the updated S(
, 
) of JENDL-5 might affect the result of critical characteristics in the critical analyses for the modified STACY core configuration.
