Faculty Profiles - ENDO Tomohiro

Papers - ENDO Tomohiro

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Verification of In-core power distribution reconstruction based on POD for high-temperature gas-cooled reactors using ex-core detectors Reviewed International journal

Yuki Urase, Akio Yamamoto, Tomohiro Endo

Journal of Nuclear Science and Technology Vol. 63 ( 1 ) page： 66 - 81 2026.1

Language：English Publishing type：Research paper (scientific journal) Publisher：Talylor & Francis

A reconstruction method of in-core power distribution using the proper orthogonal decomposition (POD) and the signal from ex-core detectors is proposed. POD can efficiently reconstruct the in-core power distribution from a small number of constraints (i.e. detectors) by expressing the power (≈ neutron flux) distribution in terms of a small number of POD bases and expansion coefficients. The number of unknowns required to reconstruct the power distribution is reduced from the number of spatial meshes in the core calculation to the number of POD expansion coefficients. A new method using the linear regression is proposed to calculate the POD expansion coefficients in addition to the method based on the generalized inverse matrix. In this study, the applicability of the in-core power distribution reconstruction method based on the ex-core detector signals and POD is verified in a two-dimensional system simulating the High-Temperature Engineering Test Reactor (HTTR). As a result of the verification, it was confirmed that the in-core power distribution can be reconstructed with a limited number of ex-core detector signals. The influence of statistical errors in the detector signals was also evaluated, and the in-core power distribution can be appropriately reconstructed when the order of the detector counts is approximately less than 3%. The linear regression method was numerically more stable than the generalized inverse matrix method in calculating POD expansion coefficients.

DOI： 10.1080/00223131.2025.2533530

Fundamental properties and characteristics of flux distribution tallies using proper orthogonal decomposition Reviewed International journal

Ryoichi Kondo, Akio Yamamoto, Tomohiro Endo

EPJ Nuclear Sciences & Technologies Vol. 11 page： 21 2025.6

Authorship：Last author Language：English Publishing type：Research paper (scientific journal) Publisher：EDP Sciences

The flux distribution tallies using the proper orthogonal decomposition (POD) called “the POD tallies” have been developed in our previous study. The POD tallies can achieve dimensionality and statistical uncertainty reduction for a finely discretized flux distribution. Some characteristics of the POD tallies, which are left by our previous work, are revealed in the present study. Firstly, the POD tallies with the track length estimator are newly implemented. Since the statistical uncertainty of the POD tallies is reduced compared with the cell tallies, the POD tallies with the track length estimator can obtain the most precise result among the present implantations. Secondly, the basis vectors obtained by the deterministic and the stochastic methods are compared. The statistical uncertainty of the snapshot data invokes the degradation of the extracted basis vectors. This result indicates that the deterministic method might be more efficient for the snapshot calculation. Finally, the impact of the covariances of expansion coefficients on the statistical uncertainty of expanded flux distribution is investigated. The reconstructed statistical uncertainty considering only the variances of the expansion coefficients differs from the reference. This result reveals that the covariances of the expansion coefficients are important to estimate the statistical uncertainty of the local flux in the flux distribution.

DOI： 10.1051/epjn/2025019

Reduction of γ-ray-induced Noise of Diamond Detector Elements and Estimation of Neutron Detection Efficiency for the Development of a Criticality Proximity Monitoring System for the Decommissioning of the Fukushima Daiichi Nuclear Power Plant Reviewed International journal

Kengo Oda, Junichi H. Kaneko, Yusuke Kobayakawa, Kenichi Watanabe, Youichi Fujita, Eitaro Hamada, Tetsuichi Kishishita, Masaya Miyahara, Masayoshi Shoji, Hironori Uchinoyae, Sho Nishino, Yoshihiko Tanimura, Akiyoshi Chayahara, Takehiro Shimaoka, Hideaki Yamada, Tomohiro Endo, Yuto Akashi, Manobu M. Tanaka

Sensors and Materials Vol. 37 ( 5 ) page： 1977 - 1991 2025.5

Language：English Publishing type：Research paper (scientific journal) Publisher：MYU K.K.

In this paper, we discuss the development of a diamond detector element for a criticality proximity monitoring system, which is essential for the decommissioning of the Fukushima Daiichi Nuclear Power Plant. The system must withstand γ-rays at a dose rate of 1 kGy/h with minimal shielding due to payload constraints. We evaluated the effects of γ-rays on prototype diamond detectors and confirmed that removing the lift-off separation surface of diamond membrane by ion beam etching and adding a p+ diamond layer effectively reduced the effect of γ-rays. A prototype detector combined a 180-µm-thick 6LiF sintered neutron-to-charged-particle converter with a diamond detector having a 2.53 mm² sensitive area. With a threshold energy set at 1 MeV, a neutron detection efficiency of 3.0 × 10−4 cps/nv was obtained for a 252Cf source. The combination of a radiation-resistant front-end integrated circuit and a diamond detector resulted in an estimated γ-ray noise level of 0.001 cps at 0.915 MeV, based on the measurements with γ-rays at a dose rate of up to 600 Gy/h using a 60Co source. Furthermore, at a dose rate of 1 kGy/h, the γ-ray noise was 0.0004 cps, satisfying the S/N ratio of ≥1 required by the Feynman-α method. We suggest that by using 1024 diamonds in the future, the criticality proximity monitoring system could achieve a neutron detection efficiency of 1.9 cps/nv.

DOI： 10.18494/sam5597

Development of Multigroup Monte Carlo Neutron Transport Method with Regionwise Even-Parity Discontinuity Factor Reviewed International journal

Yoshiki Oshima, Tomohiro Endo, Akio Yamamoto

Nuclear Science and Engineering Vol. 199 ( 4 ) page： 586 - 598 2025.4

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis

The multigroup Monte Carlo (MC) neutron transport method with a regionwise even-parity discontinuity factor (REPDF), i.e. the discontinuity factor (DF)–MC method, is developed with the aim to provide a reference solution for deterministic transport calculations with DF. Applying the analogy with optics, neutrons are transmitted or reflected at a region surface during random walks. The probability of transmission or reflection is determined by REPDFs in adjacent regions. The DF is traditionally used in deterministic neutron transport methods to reduce the discretization error due to spatial homogenization and energy condensation. The DF-MC method can treat DF in the framework of the multigroup MC method.
In this paper, the weight cancellation technique based on the closest pair of points using the divide-and-conquer algorithm is used because negative weights appear due to the neutron reflection. The REPDF is calculated by the method of characteristics (MOC). The verification calculations are carried out in the pin-by-pin homogenized and assembly homogenized KAIST-2A core geometry. The DF-MC calculation can reproduce the results of the MOC with the REPDF. These results demonstrate the principle of the DF-MC method and extend the application of the DF to the probabilistic neutron transport method.

DOI： 10.1080/00295639.2024.2383102

Reduced order SP3 core calculation based on Local/Global iterations using proper orthogonal decomposition Reviewed International journal Open Access

Masato Ito, Tomohiro Endo, Akio Yamamoto, Taichi Takeishi, Yasuhiro Kodama, Hideaki Hyoudou

Journal of Nuclear Science and Technology Vol. 62 ( 4 ) page： 366 - 378 2025.4

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis

For an efficient heterogeneous neutron transport calculation, this study newly proposes a reduced order core calculation method by applying proper orthogonal decomposition (POD) to the simplified P3 (SP3) calculation with the local/global iterations. The fine mesh SP3 calculation (local calculation) for each single assembly can be efficiently solved using POD, thanks to the matrix form-based SP3 equation and the relatively small system of linear equations. By reconstructing the fine mesh distribution of neutron flux and partial neutron currents from the POD bases and the expansion coefficients estimated by the local calculation, the partial current-based coarse mesh finite difference (p-CMFD) calculation is applied to the coarse mesh calculation (global calculation). Using the global calculation results, each local boundary condition can be updated. These local and global calculations are iterated until convergence. Through a two-dimensional UO2-MOX whole core calculation with different fuel loading patterns, the proposed method is verified in comparison with the reference result by the conventional fine mesh SP3 calculation. The calculation results show that the proposed method accurately reconstructs the fine mesh distribution of neutron flux.

DOI： 10.1080/00223131.2024.2436958

Open Access

Flux distribution tallies using proper orthogonal decomposition in Monte Carlo calculations Reviewed International journal

Ryoichi Kondo, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology Vol. 61 ( 12 ) page： 1536 - 1545 2024.12

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis

In this paper, a new tallying method for a neutron flux distribution using the proper orthogonal decomposition is proposed for dimensionality reduction. The target spatial flux distribution is expanded by orthogonal basis vectors. Expansion coefficients are tallied during the random walk of the Monte Carlo calculation. The orthogonal basis vectors are extracted from the pre-calculated snapshots by the singular value decomposition. The proposed method is verified in the multi-group Monte Carlo calculation with the one-dimensional heterogeneous whole core geometry as a feasibility study. The flux distribution for each of the assemblies and energy groups is expanded by the basis vectors. The fewer basis vectors obtained from snapshots can reconstruct the target distribution well compared with the conventional Legendre polynomials used in the functional expansion tallies. The dimension of the solution in the proposed method is reduced by a factor of twenty compared with the conventional cell tally. In addition, the statistical error is reduced through dimensionality reduction thanks to the methodological feature of the proposed method. The results indicate that the proposed method has the capability of dimensionality reduction to tally the finely discretized flux distribution.

DOI： 10.1080/00223131.2024.2365445

Limited linear source approximation with Edge Detection for Convergence Stability of Method of Characteristics Reviewed International journal

Akio Yamamoto, Tomohiro Endo, Go Chiba

Journal of Nuclear Science and Technology Vol. 61 ( 11 ) page： 1405 - 1414 2024.11

Language：English Publishing type：Research paper (scientific journal) Publisher：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

Quantifying uncertainty induced by scattering angle distribution using maximum entropy method Reviewed International journal

Shuhei Maruyama, Akio Yamamoto, Tomohiro Endo

Annals of Nuclear Energy Vol. 205 page： 110591 2024.9

Authorship：Last author Language：English Publishing type：Research paper (scientific journal) Publisher：Elsevier B.V.

This study developed a new method for evaluating the uncertainty in reactor core/shielding characteristics attributable to the scattering angle distribution in the combined method of the continuous energy Monte Carlo (CEMC) transport calculations and the random sampling (RS) technique based on nuclear data covariances. Recent advances in computer performance have made it possible to evaluate the nuclear data-induced uncertainty with high accuracy using the Monte Carlo method. The total Monte Carlo (TMC) method is a typical uncertainty quantification method that relies entirely on the Monte Carlo method. Another representative Monte Carlo-based uncertainty quantification method is the combination method of the RS technique based on nuclear data covariances and CEMC transport calculations (CEMC-RS). CEMC-RS has the advantage that the uncertainty can be quantified even if one is not familiar with nuclear data measurement and evaluation, as long as the nuclear data covariances are available. However, no method has been established to quantify the uncertainty in CEMC-RS, despite the fact that the uncertainty due to the scattering angle distribution is non-negligible in fast reactor core analysis and shielding analysis. This study introduces a new approach for uncertainty quantification related to scattering angle distribution in CEMC-RS, utilizing the maximum entropy method. The effectiveness of this method was verified through comparison with results from the classical deterministic uncertainty quantification approach based on generalized perturbation theory. Overall, this method offers a more accurate tool for nuclear engineers and researchers in evaluating and managing uncertainties in reactor design and safety analysis.

DOI： 10.1016/j.anucene.2024.110591

Impact of Uncertainty Reduction on Lead-Bismuth Coolant in Accelerator-Driven System Using Sample Reactivity Experiments Reviewed International journal

Ryota Katano, Akito Oizumi, Masahiro Fukushima, Cheol Ho Pyeon, Akio Yamamoto, Tomohiro Endo

Nuclear Science and Engineering Vol. 198 ( 6 ) page： 1215 - 1234 2024.6

Authorship：Last author Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

In this study, we have demonstrated that data assimilation (DA) 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 (ADSs) derived from inelastic scattering cross sections of lead and bismuth. We reevaluated and highlighted the experimental uncertainties and correlations of the sample reactivities for the DA formula. We used the MCNP6.2 code to evaluate the sample reactivities and their uncertainties and performed DA 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 ADS 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.

DOI： 10.1080/00295639.2023.2246779

Development of nuclear data processing code FRENDY version 2 Reviewed International journal

Kenichi Tada, Akio Yamamoto, Satoshi Kunieda, Chikara Konno, Ryoichi Kondo, Tomohiro Endo, Go Chiba, Michitaka Ono, Masayuki Tojo

Journal of Nuclear Science and Technology Vol. 61 ( 6 ) page： 830 - 839 2024.6

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

Nuclear data processing is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g., neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, adaptive setting of the background cross sections, consideration of the resonance upscattering, ACE file perturbation, statistical uncertainty quantification of probability table, and modification of ENDF-6 formatted file. FRENDY version 2 was released including these new functions. It generates GENDF and MATXS formatted neutron multi-group cross section files from an ACE formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code.

DOI： 10.1080/00223131.2023.2278600

Deterministic Transport Calculation Method for Statistical Geometry with Small Fuel Particles Reviewed International journal

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

Nuclear Science and Engineering Vol. 198 ( 5 ) page： 981 - 992 2024.5

Language：English Publishing type：Research paper (scientific journal) Publisher：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

Application of neutron current method for Dancoff factor estimation of fuel particles in double-heterogeneous fuel Reviewed International journal

Akio Yamamoto, Tomohiro Endo

Journal of Nuclear Science and Technology Vol. 61 ( 3 ) page： 354 - 362 2024.3

Language：English Publishing type：Research paper (scientific journal) Publisher：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 Reviewed International journal Open Access

Shuhei Maruyama, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology Vol. 61 ( 1 ) page： 31 - 43 2024.1

Language：English Publishing type：Research paper (scientific journal) Publisher：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 Reviewed International journal Open Access

Yuki Morishita, Akio Yamamoto, Tomohiro Endo

Journal of Nuclear Science and Technology Vol. 60 ( 12 ) page： 1573 - 1585 2023.12

Language：English Publishing type：Research paper (scientific journal) Publisher：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

Open Access

An estimation method for an unknown covariance in cross-section adjustment based on unbiased and consistent estimator Reviewed International journal

Shuhei Maruyama, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology Vol. 60 ( 11 ) page： 1372 - 1385 2023.11

Language：English Publishing type：Research paper (scientific journal) Publisher：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 Reviewed International journal Open Access

Tomoaki Watanabe, Kenichi Tada, Tomohiro Endo, Akio Yamamo

Journal of Nuclear Science and Technology Vol. 60 ( 11 ) page： 1386 - 1396 2023.11

Language：English Publishing type：Research paper (scientific journal) Publisher：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 Reviewed International journal

Kenichi Tada, Tomohiro Endo

Journal of Nuclear Science and Technology Vol. 60 ( 11 ) page： 1397 - 1405 2023.11

Language：English Publishing type：Research paper (scientific journal) Publisher：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

Comparison of internal boundary conditions for optical diffusion calculations considering reflection and refraction Reviewed International journal Open Access

Toranosuke Amano, Tomohiro Endo, Akio Yamamoto

Optics Continuum Vol. 2 ( 7 ) page： 1540 - 1560 2023.6

Language：English Publishing type：Research paper (scientific journal) Publisher：Optica Publishing Group

A method to treat the internal boundary condition in an optical diffusion calculation is proposed and is compared with the conventional methods. One of the existing internal boundary conditions is Haskel's method, which uses the effective reflection coefficient for partial currents. However, Haskel's method ignores incoming partial currents from the adjacent mesh in its derivation. As a result, the accuracy at the internal boundary is lower. This paper proposes a method to improve the accuracy by iteratively updating the effective reflection coefficient for partial current. The proposed method was applied to the benchmark calculations on a one-dimensional slab geometry and its accuracy was confirmed by comparing it with the reference solution obtained by the Monte Carlo code MCML, along with the previously proposed Haskel's method and Aronson's method. As a result, it was confirmed that the proposed method is more accurate than Haskel's method at the internal boundary and gives the same result as Aronson's method. The convergence of the effective reflection coefficient using iterative calculations in the proposed method was good.

DOI： 10.1364/optcon.492445

Open Access

A Feynman-α analysis ranging over time constants of delayed neutrons based on time-sequence counting data consecutively acquired for a very long time Reviewed International journal

Sin-Ya Hohara, Tomohiro Endo, Kazuki Takahashi, Kunihiro Nakajima, Atsushi Sakon, Tadafumi Sano, Kengo Hashimoto

Journal of Nuclear Science and Technology Vol. 60 ( 6 ) page： 724 - 730 2023.6

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis

Time-sequence counting data of a neutron counter were acquired for 17 consecutive days at shutdown state of UTR-KINKI reactor and then a Feynman-α analysis over a very wide gate width range was carried out to give light on a delayed-neutron contribution of all precursors, where statistical error of the correlation amplitude Y was estimated using the bootstrap method. A first plateau of the Y could be observed in a gate width range of several hundred milliseconds. With an enlargement of gate width beyond a time constant of prompt neutrons, the Y monotonically increased above the first plateau of prompt neutrons. Enlarging gate width beyond 100 s, the Y seemed to approach a second plateau of delayed neutrons. The second plateau originated from the delayed neutrons released from the longest-life Br87. In the gate width range beyond 100 s, however, the statistical error increased sharply. The increasing error indicates a statistical limit resulting from the present measurement time. A value Yp of the first plateau of prompt neutrons was determined from another Feynman- analysis confined to a short gate width range and the Y was divided by the plateau’s value Yp to eliminate an unknown detection efficiency. The Y-to-Yp ratio was calculated using several sets of delayed-neutron group parameters for thermal fission of U235 nuclide, to validate these sets by comparing with the measured ratio to which all precursors contributed. The ratios calculated using Keepin’s 6-group and Spriggs’ 8-group parameters agreed very well with the measured ratio in a gate width range up to 100 s and consequently these parameters could be validated. However, the ratios calculated using the other sets were in a very poor agreement with the measured ratio in the gate width range.

DOI： 10.1080/00223131.2022.2138601

Development of ACE file perturbation tool using FRENDY Reviewed International journal

Kenichi Tada, Ryoichi Kondo, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology Vol. 60 ( 6 ) page： 624 - 631 2023.6

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis

The sensitivity analysis and the uncertainty quantification play an important role in improving the evaluated nuclear data library. The current computational performance enables us to perform sensitivity analysis and uncertainty quantification using continuous energy Monte Carlo calculation codes. 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 matrices. The uncertainty of k eff using the ACE file 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 eff. The ACE file perturbation tool is included in the current version of FRENDY.

DOI： 10.1080/00223131.2022.2130463

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Papers - ENDO Tomohiro