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Journal Articles

Deterministic sampling method using simplex ensemble and scaling method for efficient and robust uncertainty quantification

Endo, Tomohiro*; Maruyama, Shuhei; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 61(3), p.363 - 374, 2024/03

Uncertainty quantification (UQ) of the neutron multiplication factor is important to investigate the appropriate safety margin for a target system. Although the random sampling method is a practical and useful UQ method, a large computational cost is required to reduce the statistical error of the estimated uncertainty. Furthermore, if an input variable follows a normal distribution with a large standard deviation, the perturbed input variable by the random sampling method may become a physically inappropriate or negative value. To address these issues for the efficient and robust UQ, a modified deterministic sampling method using the simplex ensemble and the scaling method is proposed. The features of the proposed method are summarized as follows: The sample size is (r+2), where r corresponds to the effective rank of the covariance matrix between the input variables; depending on a situation of target UQ, the amounts of perturbations for the input parameters can be arbitrarily given by the scaling factor method; the scaling factor can be updated to avoid physically inappropriate in the perturbed input variables. The effectiveness of the proposed method is demonstrated through the UQ of the neutron multiplication factor due to fuel manufacturing uncertainties for a typical PWR pin-cell burnup calculation.

Journal Articles

Uncertainty reduction of sodium void reactivity using data from a sodium shielding experiment

Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 61(1), p.31 - 43, 2024/01

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.

Journal Articles

An Estimation method for an unknown covariance in cross-section adjustment based on unbiased and consistent estimator

Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 60(11), p.1372 - 1385, 2023/11

Journal Articles

Applicability evaluation of Akaike's Bayesian information criterion to covariance modeling in the cross-section adjustment method

Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*

EPJ Web of Conferences, 281, p.00008_1 - 00008_9, 2023/03

The applicability of Akaike's Bayesian Information Criterion (ABIC) to covariance modeling in the cross-section adjustment method was investigated. One of the most important things for a reliable cross-section adjustment method is giving a suitable covariance matrix. However, since we cannot know the true covariance matrix in advance, we usually estimate and assume it. To judge the goodness of the covariance matrix modeling, a metric is desirable. As a candidate for this metric, we focus on ABIC which is one of the information criteria in Bayesian inference, because the cross-section adjustment method is often discussed within the framework of Bayesian inference. In the conventional cross-section adjustment method, incorporation of the analysis model uncertainty in a covariance matrix still requires ad hoc treatment. In JAEA, the integral experimental database for fast reactors has been developed and the adjusted cross-section set ADJ2017 has been created based on this database. Many of the core characteristics in the database have been analyzed by a deterministic method. Therefore, the predicted core characteristics have non-negligible uncertainties with correlations due to some numerical approximations. However, the evaluations of the uncertainties and their correlations are still challenging issues. In addition, there would be unknown uncertainties that experimenters and analysts of reactor physics experiments could not recognize. To judge the goodness of the covariance matrix related to these uncertainties, the applicability of ABIC to the cross-section adjustment method was investigated.

Journal Articles

Development of a robust nuclear data adjustment method to outliers

Fukui, Yuhei*; Endo, Tomohiro*; Yamamoto, Akio*; Maruyama, Shuhei

EPJ Web of Conferences, 281, p.00006_1 - 00006_9, 2023/03

We developed a new nuclear data adjustment method for experimental data containing outliers. This method mitigates the effect of outliers by applying M-estimation, a type of robust estimation, to the conventional nuclear data adjustment method using sensitivity coefficients. Based on the M-estimation, we derived a weighted nuclear data adjustment formula and developed a weight calculation method. The weighted nuclear data adjustment formula was derived by weighting the function to take the extremum of the conventional nuclear data adjustment. The weighting of each nuclear characteristic is calculated from the difference between the measured and calculated values of the nuclear characteristic. This weight calculation method can evaluate the validity of each nuclear characteristic by considering correlations between nuclear characteristics using singular value decomposition. The proposed method and the conventional method were compared and verified by twin experiments. In the twin experiments, the nuclear data were adjusted using experimental data that intentionally included outliers. As a result of twin experiments, it was confirmed that the nuclear data were adjusted robustly and appropriately even with the experimental data containing outliers.

Journal Articles

Sodium-cooled Fast Reactors

Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.

Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07

This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.

Journal Articles

Boundary condition free homogenization and evaluation of its performance in fast reactor core analysis

Maruyama, Shuhei

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 10 Pages, 2022/05

This paper proposes a new homogenization method, "Boundary Condition Free Homogenization (BCFH)". The traditional homogenization method separates the core calculation and the cell (assembly) calculation by assuming a specific boundary condition or a peripheral region in the cell calculation. Nevertheless, there are ambiguities and approximation in these assumptions, and they can also cause a decline in accuracy. BCFH aims to avoid these problems and improve the accuracy in the cell calculation such as homogenization. We imposed the conditions that the physical quantities in the cell related to the reaction rate preservation is preserved for any incoming partial current, during the homogenization. That is, the response matrices of cell average (or total) flux and outgoing partial current, to be the same form between heterogeneous and homogeneous system. As a result, homogenized parameters, such as cross-sections, superhomgenization factors, and discontinuity factors, are no longer dependent on a specific boundary condition. The new homogenized parameters obtained in this way are extended from the conventional vector form to the matrix form in BCFH. To investigate the performance of BCFH, numerical tests are done for the simplified models which originates in 750MW-class sodium-cooled fast reactor with MOX fuel core in Japan. It is found that BCFH is particularly effective in evaluating control rod reactivity worth and reaction rate distribution, compared to the traditional method. We conclude that the BCFH can be a promising homogenization concept for core neutronic analysis.

JAEA Reports

Development of the unified cross-section set ADJ2017R

Yokoyama, Kenji; Maruyama, Shuhei; Taninaka, Hiroshi; Oki, Shigeo

JAEA-Data/Code 2021-019, 115 Pages, 2022/03

JAEA-Data-Code-2021-019.pdf:6.21MB
JAEA-Data-Code-2021-019-appendix(CD-ROM).zip:435.94MB

In JAEA, several versions of unified cross-section set for fast reactors have been developed so far; we have developed a new unified cross-section set ADJ2017R, which is an improved version of the unified cross-section setADJ2017 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses in reactor core design via the cross-section adjustment methodology; the values are stored in the standard database for FBR core design. In the methodology, the cross-section set is adjusted by integrating the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. ADJ2017R basically has the same performance as ADJ2017, but we conducted an additional investigation on ADJ2017 and revised the following two points. The first is to unify the evaluation method of the correlation coefficient of uncertainty caused by experiments (hereinafter referred to as the experimental correlation coefficient). Because it was found that the common uncertainty used in the evaluation of the experimental correlation coefficient was evaluated by two different methods, the experimental correlation coefficients were revised for all experimental data, and the evaluation method was unified. The second is the review of the integral experiment data used for the cross-section adjustment calculation. It was found that one of the experimental values of composition ratio after irradiation of the Am-243 sample has a problem in uncertainty evaluation because its experimental uncertainty is extremely small compared to the others. The cross-section adjustment calculation was, therefore, redone by excluding the experimental value. In the creation of ADJ2017, a total of 719 data sets were analyzed and evaluated, and eventually adopted 620 integral experimental data sets. In contrast, a total of 61

JAEA Reports

Development of the Unified Cross-section Set ADJ2017

Yokoyama, Kenji; Sugino, Kazuteru; Ishikawa, Makoto; Maruyama, Shuhei; Nagaya, Yasunobu; Numata, Kazuyuki*; Jin, Tomoyuki*

JAEA-Research 2018-011, 556 Pages, 2019/03

JAEA-Research-2018-011.pdf:19.53MB
JAEA-Research-2018-011-appendix1(DVD-ROM).zip:433.07MB
JAEA-Research-2018-011-appendix2(DVD-ROM).zip:580.12MB
JAEA-Research-2018-011-appendix3(DVD-ROM).zip:9.17MB

We have developed a new unified cross-section set ADJ2017, which is an improved version of the unified cross-section set ADJ2010 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses; the values are stored in the standard database for FBR core design via the cross-section adjustment methodology, which integrates with the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. The ADJ2017 is based on Japan's latest nuclear data library JENDL-4.0 as in the previous version of ADJ2010, and it incorporates more information on integral experimental data sets related to minor actinides (MAs) and degraded plutonium (Pu). In the creation of ADJ2010, a total of 643 integral experimental data sets were analyzed and evaluated, and 488 of the integral experimental data sets were finally selected to be used for the cross-section adjustment. In contrast, we have evaluated a total of 719 data sets, and eventually adopted 620 integral experimental data sets to create ADJ2017. ADJ2017 shows almost the same performance as ADJ2010 for the main neutronic characteristics of conventional sodium-cooled MOX-fuel fast reactors. In addition, for the neutronic characteristics related to MA and degraded Pu, ADJ2017 improves the C/E values of the integral experimental data sets, and reduces the uncertainty induced by the nuclear data. ADJ2017 is expected to be widely used in the analysis and design research of fast reactors. Moreover, it is expected that the integral experimental data sets used for ADJ2017 can be utilized as a standard database of FBR core design.

Journal Articles

U-Pb dating of calcite using LA-ICP-MS; Instrumental setup for non-matrix-matched age dating and determination of analytical areas using elemental imaging

Yokoyama, Tatsunori; Kimura, Junichi*; Mitsuguchi, Takehiro; Danhara, Toru*; Hirata, Takafumi*; Sakata, Shuhei*; Iwano, Hideki*; Maruyama, Seiji*; Chang, Q.*; Miyazaki, Takashi*; et al.

Geochemical Journal, 52(6), p.531 - 540, 2018/12

 Times Cited Count:16 Percentile:64.36(Geochemistry & Geophysics)

Journal Articles

Irradiation induced reactivity in Monju zero power operation

Takano, Kazuya; Maruyama, Shuhei; Hazama, Taira; Usami, Shin

Proceedings of Reactor Physics Paving the Way Towards More Efficient Systems (PHYSOR 2018) (USB Flash Drive), p.1725 - 1735, 2018/04

Irradiation dependence of the core excess reactivity was investigated for the Monju system startup tests at zero-power carried out in 2010. The excess reactivity basically decreases with the $$beta$$ decay of $$^{241}$$Pu in zero-power operation. However, the excess reactivity little changed in the two month period of the startup tests, which suggests a positive reactivity insertion during the period. The investigated irradiation dependence shows that the positive reactivity increases with reactor operation and mostly saturates by the fission-dose attained during the Monju zero-power operation in a month ($$sim$$10$$^{17}$$ fissions/cm$$^{3}$$). The saturated positive reactivity is equivalent to approximately 47% of the initially accumulated self-irradiation damage recovery assuming the defects were recovered by the fission-fragment irradiation in the reactor operation.

Journal Articles

Model verification and validation procedure for a neutronics design methodology of next generation fast reactors

Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

Journal Articles

Core performance requirements and design conditions for next-generation sodium-cooled fast reactor in Japan

Oki, Shigeo; Maruyama, Shuhei; Chikazawa, Yoshitaka; Ohtaki, Akira; Kubo, Shigenobu; Hibi, Koki*; Kan, Taro*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04

Journal Articles

Core design of the next-generation sodium-cooled fast reactor in Japan

Kan, Taro*; Ogura, Masashi*; Hibi, Koki*; Oki, Shigeo; Maeda, Seiichiro; Maruyama, Shuhei; Ohgama, Kazuya

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

Journal Articles

Development of a fast reactor for minor actinides transmutation; Improvement of prediction accuracy for MA-related integral parameters based on cross-section adjustment technique

Yokoyama, Kenji; Maruyama, Shuhei; Numata, Kazuyuki; Ishikawa, Makoto; Takeda, Toshikazu*

Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.1906 - 1915, 2016/05

Journal Articles

Precise determination of $$^{12}_{Lambda}$$C level structure by $$gamma$$-ray spectroscopy

Hosomi, Kenji; Ma, Y.*; Ajimura, Shuhei*; Aoki, Kanae*; Dairaku, Seishi*; Fu, Y.*; Fujioka, Hiroyuki*; Futatsukawa, Kenta*; Imoto, Wataru*; Kakiguchi, Yutaka*; et al.

Progress of Theoretical and Experimental Physics (Internet), 2015(8), p.081D01_1 - 081D01_8, 2015/08

 Times Cited Count:14 Percentile:66.75(Physics, Multidisciplinary)

Level structure of the $$^{12}_{Lambda}$$C hypernucleus was precisely determined by means of $$gamma$$-ray spectroscopy. We identified four $$gamma$$-ray transitions via the $$^{12}$$C$$(pi^{+},K^{+}gamma)$$ reaction using a germanium detector array, Hyperball2. The spacing of the ground-state doublet $$(2^{-}, 1^{-}_{1})$$ was measured to be $$161.5pm0.3$$(stat)$$pm0.3$$ (syst)keV from the direct $$M1$$ transition. Excitation energies of the $$1^{-}_{2}$$ and $$1^{-}_{3}$$ states were measured to be $$2832pm3pm4$$, keV and $$6050pm8pm7$$, keV, respectively. The obtained level energies provide definitive references for the reaction spectroscopy of $$Lambda$$ hypernuclei.

Journal Articles

On-site background measurements for the J-PARC E56 experiment; A Search for the sterile neutrino at J-PARC MLF

Ajimura, Shuhei*; Bezerra, T. J. C.*; Chauveau, E.*; Enomoto, T.*; Furuta, Hisataka*; Harada, Masahide; Hasegawa, Shoichi; Hiraiwa, T.*; Igarashi, Yoichi*; Iwai, Eito*; et al.

Progress of Theoretical and Experimental Physics (Internet), 2015(6), p.063C01_1 - 063C01_19, 2015/06

 Times Cited Count:6 Percentile:45.45(Physics, Multidisciplinary)

The J-PARC E56 experiment aims to search for sterile neutrinos at the J-PARC Materials and Life Science Experimental Facility (MLF). In order to examine the feasibility of the experiment, we measured the background rates of different detector candidate sites, which are located at the third floor of the MLF, using a detector consisting of plastic scintillators with a fiducial mass of 500 kg. The gammas and neutrons induced by the beam as well as the backgrounds from the cosmic rays were measured, and the results are described in this article.

Journal Articles

Isothermal temperature coefficient evaluation for the Monju restart core

Mori, Tetsuya; Maruyama, Shuhei; Hazama, Taira; Suzuki, Takayuki

Nuclear Technology, 179(2), p.286 - 307, 2012/08

 Times Cited Count:11 Percentile:63.54(Nuclear Science & Technology)

The present paper describes the evaluation of the isothermal temperature coefficient data obtained in the Monju restart core. As in the preceding evaluations on the criticality and the control rod worth, the best-estimate value and its uncertainty are evaluated as accurately as possible. Data obtained in the previous test is evaluated in the same level of detail. The measured data shows that the fuel composition change from the previous test decreases the magnitude of the temperature coefficient by $$sim$$8%. Through a sensitivity analysis, it is confirmed that the decrease is mainly brought by the composition of $$^{241}$$Pu and $$^{241}$$Am. The best accuracy within the experimental uncertainty of 2% is attained for the previous core by a calculation with JENDL-4.0. Results for the restart core show inconsistent behavior and require a further investigation.

Journal Articles

Correlations among FBR core characteristics for various fuel compositions

Maruyama, Shuhei; Oki, Shigeo; Okubo, Tsutomu; Kawashima, Katsuyuki; Mizuno, Tomoyasu

Journal of Nuclear Science and Technology, 49(6), p.640 - 654, 2012/06

 Times Cited Count:3 Percentile:25.07(Nuclear Science & Technology)

This study shows the good correlations in FBR core characteristics, and find out the mechanism of their correlations with the aid of sensitivity analyses. It has been clarified that Doppler coefficient turns to have the correlations with the other core characteristics by considering the constraint of the criticality requirement for fuel composition variations. The finding of the correlations makes easy to specify the ranges of core reactivity control and core safety properties which are important for core design in determining core specification and performance. It gives significant information for FBR core design in the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. By using this index and the correlations, the core characteristic variations can be estimated for various fuel compositions without repeating core calculations.

Journal Articles

Consideration of methods to determine an enrichment of commercial fast reactor fuel

Maruyama, Shuhei; Oki, Shigeo; Okubo, Tsutomu

Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.1635 - 1643, 2011/05

In the situation that a variety of fuel composition is fed to FBRs during LWR-to-FBR transition stage, a special consideration on fuel reactivity and its loss by burnup is needed for each feeding fuel to keep criticality during operation period as prescribed. Suitable methods of determination of fuel enrichment will accomplish this without changing a pattern of fuel-loading (including number of loading fuel assembly). The choice of the method of enrichment determination affects the core characteristics which have to be controlled in core design. This paper describes some characteristics of the methods to determine an enrichment of fast reactor fuel from the core design points of view. Merits and demerits of these methods had been clarified in this study.

53 (Records 1-20 displayed on this page)