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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.
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.
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：0 Percentile：0.00(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.
Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05
Times Cited Count：1 Percentile：25.62(Nuclear Science & Technology)Fukushima, Masahiro; Okajima, Shigeaki*; Mukaiyama, Takehiko*
Journal of Nuclear Science and Technology, 61(4), p.478 - 497, 2024/04
Times Cited Count：3 Percentile：62.75(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.
Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka
Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01
Times Cited Count：0 Percentile：0.00(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：2 Percentile：46.61(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
Fukuda, Kodai
Proceedings of 4th Reactor Physics Asia Conference (RPHA2023) (Internet), 4 Pages, 2023/10
Brief evaluations were performed using the N-F model to quantitatively clarify the effect of thermal expansion on the consequences of criticality accidents in the water-moderated fuel-particle-dispersion system. The analysis clarified that ignoring thermal expansion can lead to underestimation or overestimation of the consequences by several tens of percent. It is concluded that evaluators can ignore the thermal expansion when they evaluate the consequences of the prompt supercritical transient in water-moderated solid fuel-dispersion systems, such as fuel debris systems. Only the Doppler effect can be considered when the fuel-temperature-feedback coefficient is prepared. However, depending on the required accuracy, the evaluators should take care of the error caused by ignoring thermal expansion.
Ota, Hirokazu*; Ogata, Takanari*; Yamano, Hidemasa; Futagami, Satoshi; Shimada, Sadae*; Yamada, Yumi*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05
Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa
Journal of Nuclear Engineering and Radiation Science, 9(2), p.021601_1 - 021601_9, 2023/04
Feedback reactivity automatically caused by radial expansion of the core is known as one of the inherent safety features in a sodium-cooled fast reactor (SFR). In order to validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD, the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of BOP-302R and BOP-301 in an experimental SFR, EBR-II were conducted and the applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS even, by comparing the numerical results and the experimental data.
Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04
Yamane, Yuichi
Journal of Nuclear Science and Technology, 59(11), p.1331 - 1344, 2022/11
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The reactivity was estimated from a time profile of neutron count rate or a simulated data in a quasi-steady state after sudden change of reactivity or external neutron source strength. The estimation was based on the equation of power in subcritical quasi-steady state. The purpose of the study is to develop the method of timely reactivity estimation from complicated time profile of neutron count rate. The developed method was applied to the data simulating neutron count rate created by using one-point kinetics code, AGNES, and Poisson-distributed random noise and to the transient subcritical experiment data measured by using TRACY. The result shows that the difference of the estimated and reference value was within about 5% or less for (
-1) for simulated data and within about 7% or less for 
-1.4 and -3.1 for the experimental data. It was also shown that the possibility of the reactivity estimation several ten seconds after the status change.
Ohgama, Kazuya; Takegoshi, Atsushi*; Katagiri, Hiroki; Hazama, Taira
Nuclear Technology, 208(10), p.1619 - 1633, 2022/10
Times Cited Count：4 Percentile：53.26(Nuclear Science & Technology)Katano, Ryota; Oizumi, Akito; Fukushima, Masahiro; Pyeon, C. H.*
KURNS Progress Report 2020, P. 102, 2021/07
For the design study of ADS, integral experimental data of LBE is necessary to validate cross sections of lead (Pb) and bismuth (Bi). In this study, we conducted Pb and Bi void reactivity measurements using aluminum (Al) void space in Kyoto University Critical Assembly (KUCA). We found that the calculations overestimate the void reactivities of Pb and Bi by about 20 pcm.
Kakiuchi, Kazuo; Udagawa, Yutaka; Amaya, Masaki
Annals of Nuclear Energy, 155, p.108171_1 - 108171_11, 2021/06
Times Cited Count：2 Percentile：20.79(Nuclear Science & Technology)Yamane, Yuichi
Journal of Nuclear Science and Technology, 57(8), p.926 - 931, 2020/08
Times Cited Count：1 Percentile：8.93(Nuclear Science & Technology)An equation of power in subcritical quasi-steady state has been derived based on one-point kinetics equations for the purpose of utilizing it for the development of timely reactivity estimation from complicated time profile of neutron count rate. It linearly relates power, 
, to a new variable 
, which is a function of time differential of the power. It has been confirmed by using one-point kinetics code, AGNES, that the calculated points (
) are perfectly in a line described by the new equation and that points () calculated from transient subcritical experiments by using TRACY made a line with a slope indicated by the new equation.
Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki
Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06
Times Cited Count：3 Percentile：27.12(Nuclear Science & Technology)Taniguchi, Yoshinori; Udagawa, Yutaka; Amaya, Masaki
Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05
Times Cited Count：1 Percentile：8.93(Nuclear Science & Technology)
and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki
Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05
Times Cited Count：3 Percentile：27.12(Nuclear Science & Technology)Fukushima, Masahiro; Goda, J.*; Oizumi, Akito; Bounds, J.*; Cutler, T.*; Grove, T.*; Hayes, D.*; Hutchinson, J.*; McKenzie, G.*; McSpaden, A.*; et al.
Nuclear Science and Engineering, 194(2), p.138 - 153, 2020/02
Times Cited Count：7 Percentile：53.80(Nuclear Science & Technology)To validate lead (Pb) nuclear cross sections, a series of integral experiments to measure lead void reactivity worth was conducted systematically in three fast spectra with different fuel compositions on the Comet critical assembly of the National Criticality Experiments Research Center. Previous experiments in a high-enriched uranium (HEU)/Pb and a low-enriched uranium (LEU)/Pb systems had been performed in 2016 and 2017, respectively. A follow-on experiment in a plutonium (Pu)/Pb system has been completed. The Pu/Pb system was constructed using lead plates and weapons grade plutonium plates that had been used in the Zero Power Physics Reactor (ZPPR) of Argonne National Laboratory until the 1990s. Furthermore, the HEU/Pb system was re-examined on the Comet critical assembly installed newly with a device that can guarantee the gap reproducibility with a higher accuracy and precision, and then the experimental data was re evaluated. Using the lead void reactivity worth measured in these three cores with different fuel compositions, the latest nuclear data libraries, JENDL 4.0 and ENDF/B VIII.0, were tested with the Monte Carlo calculation code MCNP version 6.1. As a result, the calculations by ENDF/B-VIII.0 were confirmed to agree with lead void reactivity worth measured in all the cores. It was furthermore found that the calculations by JENDL 4.0 overestimate by more than 20% for the Pu/Pb core while being in good agreements for the HEU/Pb and LEU/Pb cores.
