研究者詳細 - 遠藤　知弘

論文 - 遠藤　知弘

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Bias factor method using random sampling technique 査読有り国際誌

Tomohiro Endo, Akio Yamamoto, Tomoaki Watanabe

Journal of Nuclear Science and Technology 53 巻 ( 10 ) 頁： 1494 - 1501 2016年10月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis, Ltd.

Toward the practical use of the bias factor method for actual light water reactor core analyses, the bias factor method using the random sampling technique is newly proposed. The bias factor method is one of the correction methods using information of E/C values in existing measurable systems, to reduce biases and uncertainties of predicted core characteristics parameters. By the aid of the random sampling technique, our proposed bias factor method can be carried out using only forward calculations without any adjoint calculations, and can easily take into account burnup and thermal-hydraulic feedback effects, which are difficult points in the practical application to actual core analyses. Although the statistical error due to the random sampling technique is inevitable in the proposed method, the statistical error can be simply quantified by the resampling technique such as the bootstrap method. As one of the feasibility studies, effectiveness of the proposed method is verified through a numerical experiment which virtually simulates a typical equilibrium pressurized water reactor core. In this verification problem, it is clarified that E/C values of control rod worth at the beginning of cycle under the hot zero power condition are useful information to reduce biases and uncertainties of predicted assembly-wise power distributions during operation of hot full power.

DOI： 10.1080/00223131.2015.1126541

Scopus

その他リンク： https://doi.org/10.1080/00223131.2015.1126541

Application of dynamic mode decomposition to Rossi-α method in a critical state using file-by-file moving block bootstrap method 査読有り国際誌

Tomohiro Endo, Fuga Nishioka, Akio Yamamoto, Kenichi Watanabe, Cheol Ho Pyeon

Journal of Nuclear Science and Technology 59 巻 ( 9 ) 頁： 1117 - 1126 2022年9月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis

Prompt neutron decay constant α in a critical state is useful information to validate the numerically predicted ratio of the point kinetics parameters βeff/ℓ, where βeff and ℓ are the effective delayed neutron fraction and prompt neutron lifetime, respectively. To directly measure α in a target critical system, this study proposes the application of the dynamic mode decomposition (DMD) to the reactor noise analysis based on the Rossi-α method. The DMD-based Rossi-α method enables us to robustly estimate the fundamental mode component of α from the Rossi-α histograms measured using multiple neutron detectors. Furthermore, the file-by-file moving block bootstrap method is newly proposed for the statistical uncertainty quantification of α to prevent huge memory usage when the neutron count rate is high and/or the total measurement time is long. A critical experiment has been conducted at Kyoto University Critical Assembly to demonstrate the proposed method. As a result, the proposed method can uniquely determine the α value of which the statistical uncertainty is smallest. By utilizing this experimental result of α, numerical results of βeff/ℓ ratio using the continuous energy Monte Carlo code MCNP6.2 with recent nuclear data libraries, which are processed by the nuclear data processing code FRENDY, are validated.

DOI： 10.1080/00223131.2022.2030260

Theoretical Derivation of a Unique Combination Number Hidden in the Higher-Order Neutron Correlation Factors Using the Pál-Bell Equation 招待有り査読有り国際誌

Tomohiro Endo, Fuga Nishioka, Akio Yamamoto, Kenichi Watanabe, Cheol Ho Pyeon

Nuclear Science and Engineering 197 巻 ( 2 ) 頁： 176 - 188 2023年2月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis, Ltd.

The Pál-Bell equation is a backward-type master equation that describes the probability generating function (PGF) of neutron counts in a neutron multiplication system. Thanks to the Pál-Bell equation with the two-forked and the fundamental mode approximations, an analytical solution of PGF of neutron counts in a source-driven subcritical system can be theoretically derived. This theoretical derivation clarifies that the unique combination number of double factorial (2n−3)!! does exist in the ratio of the higher-order neutron correlation factors measured in a critical state even for any kind of fissile and moderator materials. Additionally, the unique combination numbers are experimentally validated for the order 3 ≤ n ≤ 6 through reactor noise measurements in actual subcritical systems. This knowledge can be used to judge whether a target system is in a deep subcritical state or to roughly estimate the magnitude of subcriticality, based on the factorial moments of the measured reactor noise in a zero-power state.

DOI： 10.1080/00295639.2022.2049992

Data Assimilation Using Subcritical Measurement of Prompt Neutron Decay Constant 招待有り査読有り国際誌

Tomohiro Endo, Akio Yamamoto

Nuclear Science and Engineering 194 巻 ( 11 ) 頁： 1089 - 1104 2020年11月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis, Ltd.

The prompt neutron decay constant α in a steady-state subcritical system can be directly measured using a reactor noise analysis method such as the Feynman-α method. To reduce the nuclear data-induced uncertainty of keff for a target system, this study investigates the applicability of data assimilation techniques, i.e., the bias factor method and the cross-section adjustment method, based on a subcritical measurement of α conducted at Kyoto University Critical Assembly (KUCA). The sensitivity coefficients of keff and α with respect to the nuclear data were efficiently estimated using a deterministic SN transport code with first-order perturbation theory. As a result, the a priori relative uncertainty of keff due to the 56-group SCALE covariance data can be reduced if there is strong correlation between the measured α and the target keff. The experimental value of α contributes to improving the nuclear data of total fission spectrum χ and total fission neutron number ν via strong correlations between χ and prompt χp and between ν and prompt νp , by utilizing the sensitivity coefficients of α with respect to prompt χp and νp.

DOI： 10.1080/00295639.2020.1720499

Web of Science

その他リンク： https://doi.org/10.1080/00295639.2020.1720499

Sensitivity Analysis of Prompt Neutron Decay Constant using Perturbation Theory 査読有り国際誌

Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 55 巻 ( 11 ) 頁： 1245 - 1254 2018年11月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis, Ltd.

Experimental results of prompt neutron decay constant α is useful information to validate numerical results of ω-eigenvalue for spatial and energetic fundamental mode. In order to accomplish the data assimilation technique using α, it is desirable to establish an efficient numerical calculation method for sensitivity coefficient analysis of α. For this purpose, the numerical calculation method using the first-order perturbation theory is investigated. A specific theoretical formula is derived to evaluate the sensitivity coefficient of α to nuclear data. The derived rigorous formula utilizes forward and adjoint eigenfunctions which consist of neutron flux and delayed neutron precursor densities. Using the prompt approximation, the derived formula can be simplified without the term involving the delayed neutron precursor densities. By calculating α using the multi-energy-group neutron transport code for an ICSBEP benchmark problem, the derived formula for sensitivity analysis using the perturbation theory is verified by comparing with the reference results using the direct method. Consequently, the efficient numerical procedures for uncertainty quantification of α can be established by the aid of the sensitivity coefficients based on the perturbation theory.

DOI： 10.1080/00223131.2018.1491902

Comparison of theoretical formulae and bootstrap method for statistical error estimation of Feynman-α method 査読有り国際誌

Tomohiro Endo, Akio Yamamoto

Annals of Nuclear Energy 124 巻頁： 606 - 615 2019年2月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Elsevier B.V.

This paper discusses the statistical error of the variance-to-mean ratio, or the Y value in the Feynman-α method, from a single measurement of reactor noise. As a theoretical approach, two practical theoretical formulae are derived to estimate the statistical error of Y: one is based on the propagation of uncertainty with unbiased estimators for the third-and fourth-order central moments; the other is a simplified formula that reuses the Y value under the fundamental mode approximation, where the subcriticality is approximately less than 10,000 pcm. As a numerical approach, the bootstrap method is improved to efficiently estimate the correlations of Y between different counting gate widths, or covariance matrix Σ_Y , due to the bunching method. Through an actual reactor noise experiment at the Kyoto University Criticality Assembly, the statistical errors of Y using the theoretical formulae and the bootstrap method are validated by comparing the reference statistical errors that are estimated from the multiple experiments of reactor noise. Furthermore, the impact of Σ_Y on the statistical error of the prompt neutron decay constant α is numerically investigated. Consequently, in the case of this experimental analysis, it was confirmed that the bootstrap method with the correlations of Y seems to be slightly better from the viewpoint of the coverage probability of the estimated confidence intervals of α, although the fitting error method without the correlation of Y could be useful for the order estimation of the statistical error of α.

DOI： 10.1016/j.anucene.2018.10.032

Web of Science

その他リンク： https://doi.org/10.1016/j.anucene.2018.10.032

Deterministic sampling method using simplex ensemble and scaling method for efficient and robust uncertainty quantification 査読有り国際誌

Tomohiro Endo, Shuhei Maruyama, Akio Yamamoto

Journal of Nuclear Science and Technology 61 巻 ( 3 ) 頁： 363 - 374 2024年3月

担当区分：筆頭著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis

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.

DOI： 10.1080/00223131.2023.2231931

Experimental validation of unique combination numbers for third- and fourth-order neutron correlation factors of zero-power reactor noise 査読有り国際誌

Tomohiro Endo, Akio Yamamoto, Masao Yamanaka, Cheol Ho Pyeon

Journal of Nuclear Science and Technology 56 巻 ( 4 ) 頁： 322 - 336 2019年4月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis, Ltd.

Zero-power reactor noise is useful for subcriticality measurements. Based on the nuclear reactor physics and the theory of neutron detection, this paper theoretically clarifies that the third- and fourth-order neutron correlation factors Y_3 and Y_4 can be expressed as functions of the second-order neutron correlation factor Y. In particular, if the neutron-counting gate width is sufficiently large, the saturation values Y_3⁄Y^2 and Y_4⁄Y^3 are almost equal to the unique combination numbers, ‘3’ and ‘15,’ for a source-driven subcritical system, where the subcriticality is less than 10000 pcm. These unique combination numbers, ‘3’ and ‘15,’ for Y_3⁄Y^2 and Y_4⁄Y^3 were validated using actual zero-power reactor noise measurements carried out at the Kyoto University Criticality Assembly. In this study, the estimation of statistical errors and correlations between different gate widths owing to the bunching method was achieved by the moving block bootstrap method. For a sufficiently long measured reactor noise in a steady and unperturbed state, a statistical test for the evaluation of the critical state and the absolute measurement of subcriticality can be carried out by statistically quantifying the difference between the measurement value of Y_3⁄Y^2 and the unique combination number.

DOI： 10.1080/00223131.2019.1580625

Web of Science

その他リンク： https://doi.org/10.1080/00223131.2019.1580625

Confidence interval estimation by bootstrap method for uncertainty quantification using random sampling method 招待有り査読有り国際誌

Tomohiro Endo, Tomoaki Watanabe, Akio Yamamoto

Journal of Nuclear Science and Technology 52 巻 ( 7-8 ) 頁： 993 - 999 2015年8月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis, Ltd.

Confidence interval estimation by the bootstrap method is investigated for the uncertainty quantification of neutronics calculation using the random sampling method. The random sampling method is a simple and practical technique to quantify an uncertainty (standard deviation) of the target parameter calculated by a core analysis code. It is noted that a statistical error is inevitably included in the estimated uncertainty because of the probabilistic method using random numbers. In order to estimate the statistical error of uncertainty, we focus on the bootstrap method. The bootstrap method is one of the resampling techniques to evaluate variance and confidence interval of a sample estimate (e.g. variance) without the assumption of normality. Through a lattice burnup calculation for a simplified boiling water reactor (BWR) fuel assembly, it is verified that the bootstrap method can reasonably estimate the confidence interval of uncertainty of infinite neutron multiplication factor (k(inf)) due to covariance data of JENDL-4.0. In the case of this problem, the distribution of k(inf) is well approximated by a normal distribution; thus, the confidence interval of uncertainty can be also estimated by the aid of chi-squared distribution. The merit using the bootstrap method is to simply estimate the confidence interval of uncertainty without the assumption of normality.

DOI： 10.1080/00223131.2015.1034216

Web of Science

Detected-neutron multiplication factor measured by neutron source multiplication method 査読有り国際誌

Tomohiro Endo, Akio Yamamoto, Yoshihiro Yamane

Annals of Nuclear Energy 38 巻 ( 11 ) 頁： 2417 - 2427 2011年11月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Elsevier B.V.

An alternative definition of neutron multiplication factor measured by the neutron source multiplication (NSM) method is newly proposed. This newly defined neutron multiplication factor, kdet, is derived on the basis of neutron detection process in a subcritical system with an external neutron source. The definition of kdet is expressed as a ratio of total number of detected fission-neutrons to total number of detected all neutrons. In this paper, a heuristic derivation of kdet is presented, and another interpretation of kdet is explained by using the detector importance function. Based on the idea of kdet, the measurement principle of NSM method is reinterpreted, and the correction factors in the NSM method are clarified. In order to verify our proposed NSM method, numerical analysis of the NSM method is carried out. The numerical results suggest that target neutron multiplication factors of the NSM method can be well estimated even without any corrections by putting a neutron detector where the effective neutron multiplication factor keff is well approximated by kdet.

DOI： 10.1016/j.anucene.2011.01.007

Scopus

Development of new solid angle quadrature sets to satisfy even- and odd-moment conditions 査読有り国際誌

Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 44 巻 ( 10 ) 頁： 1249 - 1258 2007年10月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis, Ltd.

In this paper, we propose new solid angle quadrature sets named EON for the SN method. The EON quadrature sets are developed for the 3-D xyz geometry, and satisfy not only even-moment conditions but also odd-moment conditions for direction cosines over an octant of the sphere. We can accurately calculate the numerical integration of the polynomial of direction cosines using EON. We verify the effectiveness of EON through the 2-D bulk shield problem and the 3-D neutron transport benchmark problems. The verification results indicate that the angular discretization error of EON is much smaller than that of the conventional level symmetric quadrature set.

DOI： 10.1080/18811248.2007.9711368

Scopus

Nuclear education programs with reactor laboratory experiments at zero-powered research reactor facilities in Japan 査読有り国際誌

Cheol Ho Pyeon, Tomohiro Endo, Go Chiba, Kenichi Watanabe, Genichiro Wakabayashi

Annals of Nuclear Energy 204 巻頁： 110531 - 110531 2024年9月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Elsevier B.V.

Reactor laboratory experiments are conducted in zero-powered research reactor facilities, at the University Teaching and Research Reactor (UTR in Kindai University: UTR-KINKI) and at Kyoto University Critical Assembly (KUCA), together with nuclear education through reactor physics experiments for university students. The experimental programs designed for undergraduate and graduate students at UTR-KINKI are classified into basic, intermediate, and advanced levels according to student coursework in subjects related to reactor physics and radiation detection; also developed is a curriculum of experimental programs to meet student needs. A new educational program is provided at KUCA, covering exponential experiments and uranium enrichment measurements. The range of the systematic development of programs for reactor laboratory experiments is extensive, covering reactor physics experiments and radiation detection. Notably, incorporating subcriticality measurements by the Feynman-α method in the UTR-KINKI core into the program offers outstanding opportunities for direct experience with measurement and data processing methods that involve critical approach monitoring of fuel debris. Introducing subcritical measurements is conducive to the fostering experts with an interest in criticality safety.

DOI： 10.1016/j.anucene.2024.110531

Deterministic Transport Calculation Method for Statistical Geometry with Small Fuel Particles 査読有り国際誌

Akio Yamamoto, Tomohiro Endo, Satoshi Takeda, Hiroki Koike, Kazuya Yamaji, Koji Asano

Nuclear Science and Engineering 198 巻 ( 5 ) 頁： 981 - 992 2024年5月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis, Ltd.

A deterministic transport calculation method is proposed for the treatment of dispersed fuel particles in a fuel compact/fuel pebble of a typical high-temperature gas-cooled reactor fuel. The random distribution of fuel particles was considered using the statistical geometry (STG) method, which is widely used in the Monte Carlo method. A long-ray trace, which represents a neutron flight path, was considered, and the segment lengths and material distributions on the ray trace were randomly sampled using STG. Then a conventional transport sweep, as used in the method of characteristics, was performed along the ray trace. The proposed deterministic statistical geometry (DSTG) method can calculate the flux spatial distribution in a heterogeneous geometry containing randomly dispersed fuel particles and the surrounding graphite matrix, which is consistent with the STG in a Monte Carlo method. The validity of the DSTG method was confirmed through sensitivity calculations and comparisons with a multigroup Monte Carlo method that utilizes STG. The proposed method can be used for the homogenization of heterogeneous structures inside a fuel compact or fuel pebble as an alternative to conventional deterministic unit cell calculations that consider fuel particles and the surrounding matrix in high-temperature gas-cooled reactor fuels.

DOI： 10.1080/00295639.2023.2230414

Limited linear source approximation with Edge Detection for Convergence Stability of Method of Characteristics 招待有り査読有り国際誌

Akio YAMAMOTO, Tomohiro ENDO, Go CHIBA

Journal of Nuclear Science and Technology xx 巻 ( xx ) 頁： xxx - xxx 2024年4月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis

A new implementation of the limited linear source approximation (LLSA) is proposed. The LLSA was previously proposed to eliminate local negative source in flux regions to mitigate numerical instability of the method of characteristics (MOC) with linear source approximation (LSA). In the present LLSA implementation, the convex edges of flux regions are used to check the local negative source to decrease the computational load. The present method is implemented in the transport code GENESIS and its effectiveness is verified through the two dimensional C5G7 benchmark problem and the simplified two-dimensional high-temperature engineering test reactor core. The calculation results indicate that the present LLSA implementation efficiently mitigates the numerical instability of MOC with LSA. Additional computational time is less than 1% of total computation time.

DOI： 10.1080/00223131.2024.2341776

Application of neutron current method for Dancoff factor estimation of fuel particles in double-heterogeneous fuel 査読有り国際誌

Akio Yamamoto, Tomohiro Endo

Journal of Nuclear Science and Technology 61 巻 ( 3 ) 頁： 354 - 362 2024年3月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis

An evaluation method of Dancoff factors of fuel particles in a typical fuel element of a high-temperature gas-cooled reactor (HTGR) is proposed based on the neutron current method that is widely used for lattice physics calculations of light water reactors. The first level of heterogeneity, i.e. dispersed fuel particles in graphite matrix, is treated by the deterministic statistical geometry (DSTG) method. The homogenized cross-sections of dispersed fuel particles are provided to the second level of heterogeneity, i.e. fuel compacts in a fuel element or a fuel pebble. The validity of the present method is confirmed through the comparison with the reference Dancoff factor obtained by the Monte Carlo and the neutron current methods, which explicitly treats the double heterogeneity. The comparison is carried out for a fuel element simulating the High Temperature Engineering Test Reactor (HTTR), which adopts a typical prismatic fuel element. The numerical results indicate that the present method well reproduces the reference Dancoff factor under various packing fractions. Since the present method can handle flexible geometry and its computation time is short, the present method will be a candidate for the Dancoff factor evaluation method in design calculations of HTGR.

DOI： 10.1080/00223131.2023.2231462

Uncertainty reduction of sodium void reactivity using data from a sodium shielding experiment 査読有り国際誌

Shuhei Maruyama, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 61 巻 ( 1 ) 頁： 31 - 43 2024年1月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis

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.

DOI： 10.1080/00223131.2023.2244512

Sensitivity analysis of risk assessment for continuous Markov process Monte Carlo method using correlated sampling method 査読有り国際誌

Yuki Morishita, Akio Yamamoto, Tomohiro Endo

Journal of Nuclear Science and Technology 60 巻 ( 12 ) 頁： 1573 - 1585 2023年12月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis

The correlated sampling method is applied to the continuous Markov-chain Monte-Carlo (CMMC) method to efficiently perform sensitivity analysis of input parameters such as the failure rate of safety components. In the correlated sampling method, the original and the perturbed samples are assumed to trace an identical accident sequence, but the weight of the perturbed sample is adjusted to incorporate the variation of input data. The present method is applied to the sensitivity analysis of the safety evaluation of spent fuel pools. The result indicates that the sensitivity analysis for the CMMC coupling method can be efficiently carried out using the correlated sampling method.

DOI： 10.1080/00223131.2023.2231464

An estimation method for an unknown covariance in cross-section adjustment based on unbiased and consistent estimator 査読有り国際誌

Shuhei Maruyama, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 60 巻 ( 11 ) 頁： 1372 - 1385 2023年11月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis

A new estimation method of an unknown covariance in the cross-section adjustment method for the development of an application library is proposed. The unknown covariance is defined by the difference between the true covariance (the population covariance) and a prior covariance assumed by an analyst. The unknown covariance is estimated using an empirical covariance consistent with the observed data. To estimate the unknown covariance, an unbiased and consistent estimator in regression analysis has been incorporated into the conventional cross-section adjustment. This estimator does not require assumptions for the probability distribution of the observation data. The statistical properties of this estimator were numerically verified. In addition, the effectiveness of the proposed method was confirmed by another numerical test using actual integral experimental data. In the second numerical test, the modeling uncertainty (covariance) due to the deterministic analysis method was assumed to be unknown. The results showed that the proposed method could practically estimate the unknown covariance and adjusted cross-sections using only prior information on covariances.

DOI： 10.1080/00223131.2023.2203707

Impact of nuclear data revised from JENDL-4.0 to JENDL-5 on PWR spent fuel nuclide composition 査読有り国際誌

Tomoaki Watanabe, Kenichi Tada, Tomohiro Endo, Akio Yamamo

Journal of Nuclear Science and Technology 60 巻 ( 11 ) 頁： 1386 - 1396 2023年11月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis

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 H2O affected the nuclide compositions of PWR spent fuels.

DOI： 10.1080/00223131.2023.2201603

Convergence behavior of statistical uncertainty in probability table for cross section in unresolved resonance region 査読有り国際誌

Kenichi Tada, Tomohiro Endo

Journal of Nuclear Science and Technology 60 巻 ( 11 ) 頁： 1397 - 1405 2023年11月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis

The probability table method is a well-known method for addressing self-shielding effects in the unresolved resonance region. A long computational time is required to generate the table. The effective way to reduce the generation time is the reduction of the number of ladders. The purpose of this study is the estimation of the optimal number of ladders using the statistical uncertainty in the probability table. To this end, the statistical uncertainty quantification method of the probability table was developed and the convergence behavior of the statistical uncertainty was investigated. The product of the probability and the average cross section was considered the target of the statistical uncertainty. The convergence behavior was investigated by using Sr-90, U-235, and U-238 from JENDL-4.0. The calculation results revealed that the convergence rate of the statistical uncertainty was different in each nuclide and it was affected by the average level spacing and the average reduced neutron width. The generation time of the probability table was less than half when the input parameter for the iteration condition was changed from the number of ladders to the tolerance value. The developed method leads to the efficient generation of probability tables and the reduction of processing time.

DOI： 10.1080/00223131.2023.2204102

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