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Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 61(1), p.31 - 43, 2024/01
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)This study investigated the feasibility of reducing the uncertainty associated with fast-reactor-core design by sharing an experimental database between different fields (e.g., reactor physics and radiation shielding) using data assimilation techniques. As the first step in this study, we focused on the ORNL sodium shielding experiment and investigated the possibility of using the experimental data to reduce the uncertainty in sodium void reactivity (SVR), which is the most important safety parameter for sodium-cooled fast reactors. A sensitivity analysis based on the Generalized Perturbation Theory was performed for the sodium shielding experiment. Using the sensitivity coefficients evaluated here and those of the sodium void reactivity previously evaluated by the JAEA, we showed that sodium shielding experimental data can contribute to the uncertainty reduction of SVR by adopting the cross-section adjustment method. Based on this study, the uncertainty reduction effect is expected to be significant, especially for SVR dominated by neutron-leakage phenomena. Although new reactor physics experimental data on SVR may be difficult to obtain, the results of this study suggest that data from sodium shielding experiments can partially substitute for this role. This study demonstrated the value of the mutual use of integral experimental data in fast reactor designs.
Doda, Norihiro; Igawa, Kenichi*; Iwasaki, Takashi*; Murakami, Satoshi*; Tanaka, Masaaki
Nuclear Engineering and Design, 410, p.112377_1 - 112377_15, 2023/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)To enhance the safety of sodium-cooled fast reactors, the decay heat in the core must be removed by natural circulation even if the AC power supply to the forced circulation equipment is lost. Under natural circulation conditions, sodium flow is driven by buoyancy, and flow velocity and temperature distribution influence each other. Thus, it is difficult to evaluate the core hot spot temperature by deterministically considering the uncertainties affecting flow and heat. In this study, a statistical evaluation method is developed for the core hot spot temperature by using Monte Carlo sampling methods. The applicability of the core hotspot evaluation method was confirmed in three representative events during natural circulation decay heat removal operations in loop-type sodium-cooled fast reactors.
Tada, Kenichi; Kondo, Ryoichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 60(6), p.624 - 631, 2023/06
Times Cited Count：2 Percentile：53.91(Nuclear Science & Technology)The sensitivity analysis and the uncertainty quantification have an important role in improving the evaluated nuclear data library. The current computational performance enables us to the sensitivity analysis and uncertainty quantification using the continuous energy Monte Carlo calculation code. The ACE file perturbation tool was developed for these calculations using modules of FRENDY. This tool perturbs the microscopic cross section, the number of neutrons per fission, and the fission spectrum. The uncertainty quantification using the random sampling method is also available if the user prepares the covariance matrix. The uncertainty of the k-effective using the perturbation tool was compared to the current sensitivity analysis codes SCALE/TSUNAMI and MCNP/KSEN. The comparison results indicated that the random sampling method using this tool accurately estimates the uncertainty of k-effective.
Yamamoto, Masahiko; Horigome, Kazushi; Kuno, Takehiko
Applied Radiation and Isotopes, 190, p.110460_1 - 110460_7, 2022/12
Times Cited Count：1 Percentile：31.61(Chemistry, Inorganic & Nuclear)Gravimetric measurement of U content in UO
with ignition in the air has been investigated. The ignition temperature, ignition time and aliquot sample mass are optimized as 900
C, 60 minutes, and 1 g, respectively. The method is validated by IDMS with uncertainty estimation. The obtained result by gravimetry is 0.78236
0.00051 g/g (k=2) and agreed with IDMS value within its uncertainty. It has been found that U in UO can be measured accurately and precisely by gravimetry.
Ikeuchi, Hirotomo; Koyama, Shinichi; Osaka, Masahiko; Takano, Masahide; Nakamura, Satoshi; Onozawa, Atsushi; Sasaki, Shinji; Onishi, Takashi; Maeda, Koji; Kirishima, Akira*; et al.
JAEA-Technology 2022-021, 224 Pages, 2022/10
JAEA-Technology-2022-021.pdf:12.32MB
A set of technology, including acid dissolving, has to be established for the analysis of content of elements/nuclides in the fuel debris samples. In this project, a blind test was performed for the purpose of clarifying the current level of analytical accuracy and establishing the alternative methods in case that the insoluble residue remains. Overall composition of the simulated fuel debris (homogenized powder having a specific composition) were quantitatively determined in the four analytical institutions in Japan by using their own dissolving and analytical techniques. The merit and drawback for each technique were then evaluated, based on which a tentative flow of the analyses of fuel debris was constructed.
Iwamoto, Hiroki
JAEA-Conf 2021-001, p.83 - 87, 2022/03
We have developed a method to generate nuclear data using Gaussian process regression [1], which is one of the machine learning technique. This method generates nuclear data by treating measured data as the training data in machine learning. Since Gaussian process regression is based on nonparametric Bayesian inference, the generated nuclear data are expressed as a predictive distribution including uncertainty information. In this presentation, the basics of the Gaussian process model, some examples of the application to nuclear data generation, and other related topics will be presented. [1] H. Iwamoto, "Generation of nuclear data using Gaussian process regression", Journal of Nuclear Science and Technology, 50:8, 932-938 (2020).
Harada, Hideo
Kaku Deta Nyusu (Internet), (129), p.35 - 43, 2021/06
This paper explained the activities of the International Nuclear Data Evaluation Co-operation (WPEC)'s subgroup SG-41, which was carried out to improve the accuracy of nuclear data. In this paper, by focusing on how to deal with discrepant data, the method of deriving recommended values from multiple measurement results was reviewed. It also explained how to derive the recommended values, including exercises.
Endo, Tomohiro*; Noguchi, Akihiro*; Yamamoto, Akio*; Tada, Kenichi
Transactions of the American Nuclear Society, 124(1), p.184 - 187, 2021/06
This study confirmed that the sensitivity analysis of the alpha-eigenvalue can be carried even for non-neutron multiplication systems such as water-only systems. The preliminary results of nuclear data-induced uncertainties of alpha-eigenvalue were smaller than the differences between numerical and experimental results of alpha-eigenvalue. For further investigation, it is necessary to reconsider the experimental bias and the nuclear data-induced uncertainty in alpha-eigenvalue due to the thermal scattering law data of water.
Iwamoto, Hiroki; Meigo, Shinichiro
JPS Conference Proceedings (Internet), 33, p.011046_1 - 011046_6, 2021/03
no abstracts in English
Tanaka, Masaaki; Nakada, Kotaro*; Kudo, Yoshiro*
Nihon Kikai Gakkai-Shi, 123(1222), p.26 - 29, 2020/09
In the nuclear engineering, simulations are used in radiation, thermal hydraulic, chemical reaction, and structural fields, and the integrated fields thereof, to be applied to the design, construction and operation of nuclear facilities. This article describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements described in the "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)". And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.
Iwamoto, Hiroki
Journal of Nuclear Science and Technology, 57(8), p.932 - 938, 2020/08
Times Cited Count：13 Percentile：86.19(Nuclear Science & Technology)We present a new approach to generate nuclear data from experimental cross section data by Gaussian process regression. This paper focuses on generating proton-induced nuclide production cross sections for nickel target. Our results provide reasonable fitting curves together with their uncertainties and suggest that this approach appears to be effective in generating or evaluating the nuclear data. Besides, our results suggest that our approach could be available for experimental design in terms of reducing the generated nuclear data uncertainty.
Takamizawa, Hisashi; Nishiyama, Yutaka; Hirano, Takashi*
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 7 Pages, 2020/08
no abstracts in English
Narukawa, Takafumi
Kaku Nenryo, (54-2), P. 3, 2019/07
no abstracts in English
Yokoyama, Kenji; Kitada, Takanori*
Journal of Nuclear Science and Technology, 56(1), p.87 - 104, 2019/01
Times Cited Count：4 Percentile：41.24(Nuclear Science & Technology)no abstracts in English
Kondo, Ryoichi*; Endo, Tomohiro*; Yamamoto, Akio*; Tada, Kenichi
Proceedings of International Conference on Mathematics and Computational Methods applied to Nuclear Science and Engineering (M&C 2019) (CD-ROM), p.1493 - 1502, 2019/00
A perturbation capability of ACE formatted cross section files was developed using the modules of FRENDY. Uncertainty quantification using MCNP was carried out for the Godiva critical experiment by the RS method. We verified the results of the RS method by comparing with those obtained by the conventional sensitivity analyses. Moreover, uncertainty reduction using the bias factor method with the RS technique was applied to kinetic parameter, i.e., neutron generation time.
Stankovskiy, A.*; Iwamoto, Hiroki; 
elik, Y.*; Van den Eynde, G.*
Annals of Nuclear Energy, 120, p.207 - 218, 2018/10
Times Cited Count：8 Percentile：62.99(Nuclear Science & Technology)Propagation of high-energy (above 20-MeV) nuclear data uncertainties on the safety related neutronic responses in accelerator driven systems has been assessed. The total core power and production of radionuclides contributing to radiation source terms were focused on. The article features a method based on the Monte Carlo sampling of random nuclear data files from the covariance matrices generated from the sets of reaction cross sections obtained with model calculations of high-energy particle interactions with matter or picked up from already existing nuclear data libraries. It has been demonstrated that nuclear data uncertainties do not need to be propagated through particle transport calculations to obtain uncertainties on the responses. This advantage allowed to investigate the convergence of the sample average to the best estimate. The number of random nuclear data file sets needed to obtain reliable uncertainty on the total core power is around 300 that results in the uncertainty of 14%. The uncertainties on the concentrations of nuclides most important for the safety assessment that are accumulated in lead-bismuth eutectic during irradiation, range from 5 to 60%. Concentrations of some nuclides exemplified by Tritium converge much slower than neutron multiplicities so that several thousands of samples are needed to ensure reliable uncertainty estimates.
Tanaka, Masaaki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 14 Pages, 2018/10
A numerical simulation code MUGTHES has been developed to estimate high cycle thermal fatigue in SFRs. In development of numerical simulation code, verification, validation, and uncertainty quantification (VVUQ) are indispensable. In this study, numerical simulation at impinging jet condition in the WATLON experiment which was the water experiment of a T-junction piping system was performed for the fundamental validation. Based on the previous studies, the simplified least square version GCI method and the area validation metrics were employed as reference methods to quantify uncertainty and to measure the degree of difference between the numerical and the experimental results, respectively. Through the examinations, the potential applicability of the MUGTHES to the thermal striping phenomena was indicated and requirements of modification in the simulation was suggested in accordance with the uncertainty values.
Narukawa, Takafumi
Kaku Nenryo, (53-2), P. 5, 2018/08
no abstracts in English
Iwamoto, Hiroki; Stankovskiy, A.*; Fiorito, L.*; Van den Eynde, G.*
Journal of Nuclear Science and Technology, 55(5), p.539 - 547, 2018/05
Times Cited Count：9 Percentile：67.52(Nuclear Science & Technology)The applicability of Monte Carlo techniques, namely the Monte Carlo sensitivity method and the random-sampling method, for uncertainty quantification of the effective delayed neutron fraction 
is investigated using the continuous-energy Monte Carlo transport code, MCNP, from the perspective of statistical convergence issues. This study focuses on the nuclear data as one of the major sources of uncertainty. For validation of the calculated , a critical configuration of the VENUS-F zero-power reactor was used. It is demonstrated that Chiba's modified 
-ratio method is superior to Bretscher's prompt -ratio method in terms of reducing the statistical uncertainty in calculating not only but also its sensitivities and the uncertainty due to nuclear data. From this result and a comparison of uncertainties obtained by the Monte Carlo sensitivity method and the random-sampling method, it is shown that the Monte Carlo sensitivity method using Chiba's modified -ratio method is the most practical for uncertainty quantification of . Finally, total uncertainty due to nuclear data for the VENUS-F critical configuration is determined to be approximately 2.7% with JENDL-4.0u, which is dominated by the delayed neutron yield of 
U.
Yamamoto, Masahiko; Holland, M. K.*; Cordaro, J. V.*; Kuno, Takehiko; Surugaya, Naoki
JAEA-Technology 2017-014, 63 Pages, 2017/06
JAEA-Technology-2017-014.pdf:4.38MB
In this study, the controlled-potential coulometry has been applied as a primary method for characterizing the Pu master solutions being used as alternative source material for IDMS spikes. The coulometry system compliance with ISO12183 has been used for measurement. It has been calibrated using equipment traceable to the SI units. Plutonium standard samples have been measured to confirm the accuracy. The relative standard deviation is below 0.05%. The results agree with the reference value within 0.05%. It is found that the Pu can be precisely analyzed by the coulometry system. Then, the Pu nitrate solution, which has been purified from mixed oxide powder containing relatively high 
Pu, has been measured. The relative standard deviation is below 0.05%. The relative expanded uncertainty is less than 0.074% at the 95% confidence interval (k=2). It is indicated that coulometric assay of Pu is fit for the purpose of characterizing reference materials.
