Mar. 2022, 日本原子力学会, 技術賞特賞　　　　　　　　　　　　　　　
山本章夫;多田健一;千葉 豪

Mar. 2012, 日本原子力学会, 技術賞特賞　　　　　　　　　　　　　　　
岩本修;柴田恵一;千葉豪

Sep. 2011, 日本原子力学会炉物理部会, 炉物理部会賞　　　　　　　　　　　　　　　
千葉豪

Sep. 2006, 日本原子力学会核データ部会, 核データ部会奨励賞　　　　　　　　　　　　　　　
千葉豪

Nuclear education programs with reactor laboratory experiments at zero-powered research reactor facilities in Japan
Cheol Ho Pyeon, Tomohiro Endo, Go Chiba, Kenichi Watanabe, Genichiro Wakabayashi
Annals of Nuclear Energy, 204, 110531, 110531, Elsevier BV, Sep. 2024, [Peer-reviewed]
Scientific journal

Limited linear source approximation with Edge Detection for Convergence Stability of Method of Characteristics
Akio YAMAMOTO, Tomohiro ENDO, Go CHIBA
Journal of Nuclear Science and Technology, Informa UK Limited, 14 Apr. 2024, [Peer-reviewed]
Scientific journal

Computation time reduction of nuclear fuel burnup calculations with the predictor–corrector method using low-order model
Fuga Miyazawa, Go Chiba, Kosuke Tsujita, Shuhei Miwa
Annals of Nuclear Energy, 195, Jan. 2024, [Peer-reviewed]
Scientific journal, Special attentions should be paid on the time discretization in nuclear fuel burnup calculations for systems including gadolinia, and several efficient methods based on the predictor–corrector method have been proposed so far. There are still requirements of the further reduction of the computation time, and we propose to use a low-order model (LOM) whose computation time is much shorter than that of a full-order model (FOM) in the corrector calculations. To mitigate the error caused by the LOM adoption, correction factors for one-group quantities are devised and introduced. The proposed method is tested on the lattice physics problems under the condition that FOM is a deterministic method based on MOC and LOM is also a deterministic method based on MOC with the coarse ray-tracing condition with fewer-group cross sections. The proposed method gives results whose accuracy is almost equivalent with that of FOM, and computation time is reduced to 60–65%.

Reactor reactivity calculations with simplified-P₃ and perturbation theories
Jun-Shuang Fan, Go Chiba
Journal of Nuclear Science and Technology, Informa UK Limited, 07 Jul. 2023, [Peer-reviewed]
Scientific journal

Japanese Evaluated Nuclear Data Library version 5: JENDL-5
Osamu Iwamoto, Nobuyuki Iwamoto, Satoshi Kunieda, Futoshi Minato, Shinsuke Nakayama, Yutaka Abe, Kohsuke Tsubakihara, Shin Okumura, Chikako Ishizuka, Tadashi Yoshida, Satoshi Chiba, Naohiko Otuka, Jean-Christophe Sublet, Hiroki Iwamoto, Kazuyoshi Yamamoto, Yasunobu Nagaya, Kenichi Tada, Chikara Konno, Norihiro Matsuda, Kenji Yokoyama, Hiroshi Taninaka, Akito Oizumi, Masahiro Fukushima, Shoichiro Okita, Go Chiba, Satoshi Sato, Masayuki Ohta, Saerom Kwon
Journal of Nuclear Science and Technology, 03 Feb. 2023, [Peer-reviewed]
English, Scientific journal

ACE-FRENDY-CBZ: a new neutronics analysis sequence using multi-group neutron transport calculations
Go Chiba, Akio Yamamoto, Kenichi Tada
Journal of Nuclear Science and Technology, 60, 2, 132, 139, 2023, [Peer-reviewed]
Scientific journal, We propose a new neutronics analysis sequence using multi-group neutron transport calculations named ACE-FRENDY-CBZ. This sequence is free from uses of any application libraries; with the ACE files as the starting point, multi-group cross sections of media comprising a target system are calculated with the FRENDY code, and multi-group neutron transport calculations are performed with modules of the CBZ code system. The ACE-FRENDY-CBZ sequence was tested against the eight fast neutron systems, and good agreement in the neutron multiplication factors with the reference Monte Carlo results was obtained within 20 pcm differences in the bare systems and within 60 pcm differences in the reflected systems. It was also found that the adoption of the consistent P approximation increases the errors. In order to investigate this issue, we adopted the sub-group method to calculate spatially-dependent current-weighted total cross sections in the reflector regions, and it was suggested that the uses of the spatially-dependent cross sections with the consistent P approximation has a possibility to further improve the numerical accuracy.

Validation of LWR fuel depletion calculation module of reactor physics code system CBZ
Go Chiba, Hiroki Harada
Journal of Nuclear Science and Technology, 60, 8, 969, 979, 2023, [Peer-reviewed]
Scientific journal, This paper presents the results of validation calculations of the nuclear fuel depletion calculation module of the CBZ reactor physics code system, CBZ/Burner. Validation calculations were conducted using the post irradiation examination data obtained at Fukushima-Daini Unit 2 and at Takahama Unit 3. The nuclide number densities calculated with CBZ/Burner were compared with the measurement values, and generally good agreement was obtained. The sensitivity coefficients of the nuclide number densities with respect to nuclear data were calculated for all concerned nuclides with the depletion perturbation calculation capability of CBZ/Burner, and the nuclear data-induced uncertainties of the nuclide number densities were quantified. From the numerical results, we can conclude that the nuclear fuel depletion calculation module for LWR in the CBZ code system was successfully validated.

Level swell analysis of stagnant water pool in filtered containment venting systems
Yasunori Yamamoto, Naoto Kitahara, Fuga Miyazawa, Go Chiba, Shuichiro Miwa, Michitsugu Mori
Progress in Nuclear Energy, 155, Jan. 2023, [Peer-reviewed]
Scientific journal, A filtered containment venting system (FCVS) prevents over pressurization of containment vessel and releasing of radioactive materials during the severe accidents in nuclear power plants. During the venting process, it has been reported that the two-phase mixture level in a wet FCVS tends to swell and fluctuate. The behaviors depend on inlet/boundary conditions and physical properties of the injected gas, which vary as the accident progresses. Proper controlling and monitoring of the FCVS pool water level is crucial because it affects filtration performance including scrubbing process and thermal-hydraulic stability. In order to investigate this phenomenon, the current study proposes a set of nitrogen and steam injection experiments using a vertical pipe with a diameter of 105 mm to evaluate the effects of flow conditions and physical properties of gases. Drift flux analysis was carried out to predict the two-phase mixture water level and its fluctuations. The experimental two-phase mixture level was consistent with the values predicted by the drift flux model for nitrogen and steam injection, and the model's capability was confirmed for the system pressure ranging from atmospheric to 0.20 MPa and initial water level ranging from 0.6 to 2.6 m for both small and large diameter pipe configurations. The fluctuation amplitudes in the current experiment were smaller than those observed in experiments conducted on small-diameter pipes. The mean two-phase mixture water level increased upon pressurization of the scrubbing pool. However, it was found that the effect of pressurization on the two-phase mixture level fluctuation amplitude was negligibly small.

Implementation of Resonance Up-scattering Treatment in FRENDY Nuclear Data Processing System
Akio Yamamoto, Tomohiro Endo, Go Chiba, Kenichi Tada
Nuclear Science and Engineering, 1, 13, Informa UK Limited, 21 Jul. 2022, [Peer-reviewed]
Scientific journal

Enhancement of applicability of high-efficiency random sampling method using control variates method and sensitivity coefficients
Takumi Kida, Go Chiba
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 59, 7, 866, 874, TAYLOR & FRANCIS LTD, Jul. 2022, [Peer-reviewed]
English, Scientific journal, The CV-S method is a high-efficiency random sampling method to estimate statistical moments of random variables, and it uses an approximated target parameter which are linearly dependent on input as a mockup parameter. In order to enhance the applicability of the CV-S method, we propose to use a mockup parameter which is different from but similar to a target parameter and whose sensitivity coefficients are available. In the present work, nuclear fuel burnup problems are concerned, and standard deviation of k infinity and nuclide number densities at certain fuel burnup are estimated by the CV-S method. Through numerical tests, it is clearly demonstrated that even if sensitivity coefficients of non burnup-related parameters in a simple system like a fuel pin-cell are used as the mockup, the CV-S method has a potential to efficiently estimate statistical moments of burnup-related parameters in a complicated system like a fuel assembly.

Experimental analysis of small sample reactivity measured in the SEG experiment by a deterministic reactor physics code system CBZ
Go Chiba, Junshuang Fan
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 59, 2, 247, 256, TAYLOR & FRANCIS LTD, Feb. 2022, [Peer-reviewed]
English, Scientific journal, Experimental analysis of the sample reactivity measured in the SEG experiment is carried out with the deterministic reactor physics code system CBZ with the recent evaluated nuclear data files, JENDL-4.0, ENDF/B-VIII.0, and JEFF-3.3. Since the systems to be analysed are fast-thermal coupled ones, 211-energy group neutron reaction cross section libraries applicable to both the fast and thermal neutron systems are generated and utilized. In the multi-group library generation, the recently developed FRENDY and FRENDY/MG codes are used. Forward and adjoint neutron fluxes at the sample position are calculated by solving the neutron transport equation, and the sample reactivity is obtained by the first-order perturbation calculations. In order to simplify the systems calculated, two-dimensional cylinder model is prepared based on the previous work. Whereas the simplified model is employed, generally the reactivity of many different samples is well predicted by the calculations in comparison with the experimental uncertainties. On some of the samples, large discrepancies of the C/E values from unity are observed, and also relatively large differences in the C/E values among different nuclear data files are observed. These information are still useful for future development of the evaluated nuclear data files.

Development and verification of fast reactor burnup calculation module FRBurner in code system CBZ
Junshuang Fan, Go Chiba
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 58, 12, 1269, 1287, TAYLOR & FRANCIS LTD, Jun. 2021, [Peer-reviewed]
English, Scientific journal, CBZ is a general-purpose reactor physics analysis code system, and FRBurner, which focuses on fast reactor burnup calculations, was developed recently with diverse combinations of available methodologies. Verification of this module is conducted with the OECD/NEA fast rector benchmark since this benchmark provides various types of fast reactors. Four key reactor physics parameters, effective neutron multiplication factor k(eff), effective delayed neutron fraction beta(eff), sodium void reactivity increment rho(void), and Doppler reactivity increment rho(Doppler) are the focus and compared to two references provided by JAEA and CEA, respectively. The biases between the results from FRBurner and the JAEA and CEA references on each of the above key parameters are less than 0.5%, 1%, 3% and 7%, and less than 1.0%, 4%, 12%, and 12%, respectively. The comparison indicates that the FRBurner module would provide acceptable results for general-type fast reactor physics analysis in research. As one innovation, the detailed burnup chain model, which is significantly different from a generally used pseudo fission product model in fast reactor neutronic analysis, is applied in FRBurner. The detailed burnup chain model helps FRBurner explicitly provide information about the inventory of fission products for nuclear waste management and spent fuel reprocessing.

Improvement of the optimally-weighted predictor-corrector method for nuclear fuel burnup calculations
Jumpei Sasuga, Go Chiba, Yasunori Ohoka, Kento Yamamoto, Yasuhiro Kodama, Hiroaki Nagano
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, TAYLOR & FRANCIS LTD, Jun. 2021, [Peer-reviewed]
English, Scientific journal, A new burnup calculation method for burnable absorber-containing fuels called the Advanced Optimally-Weighted Predictor-Corrector (AOWPC) method is proposed based on the OWPC method. The AOWPC method can effectively reduce the time discretization errors in the burnup calculation. Verification calculations comparing with the conventional predictor-corrector (PC) and projected PC (PPC) methods are performed for the PWR and BWR fuel assemblies. The calculation accuracy of the AOWPC method is higher than the conventional PC method, whose burnup step is one third of the AOWPC method, and the PPC method, whose burnup step is two thirds of the AOWPC method. The additional computation time per time step for the AOWPC method is negligible compared to those for the conventional PC and PPC methods. Therefore, the computation time of the burnup calculation can be reduced using the AOWPC method.

Multi-group neutron cross section generation capability for FRENDY nuclear data processing code
Akio Yamamoto, Kenichi Tada, Go Chiba, Tomohiro Endo
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, TAYLOR & FRANCIS LTD, Jun. 2021, [Peer-reviewed]
English, Scientific journal, The multi-group cross-section generation capability for neutrons is implemented in the FRENDY nuclear data processing code. ACE-formatted files are used as the source of nuclear data instead of ENDF-formatted files since FRENDY already has the capability to generate pointwise cross sections in the ACE format. The several distinguished features are implemented for the multi-group generation capability, e.g. explicit consideration of resonance interference effect among nuclides, enhanced resonance treatment for various nuclear reactions, and accurate numerical integration of thermal cross sections. Verification calculations of the newly implemented capability are carried out through the comparison with the NJOY nuclear data processing code. Cross-section generations for all nuclides in JENDL-4.0, -4.0u, -5 alpha 4, ENDF/B-VII.1, -VIII.0, JEFF-3.3, and TENDL-2019 are carried out without unexpected processing issues, except for Pu-238 of TENDL-2019 that includes inconsistent data. The verification results indicate that the multi-group cross sections generated by FRENDY are consistent with those generated by NJOY. Now FRENDY can generate not only the pointwise cross sections for continuous energy Monte-Carlo codes but also the multi-group cross sections for deterministic neutronics analysis codes.

Sensitivity of the neutron multiplication factor to gadolinium isotopes' nuclear data for light water reactor fuel assemblies in the peak reactivity burnup range
Go Chiba
ANNALS OF NUCLEAR ENERGY, 151, PERGAMON-ELSEVIER SCIENCE LTD, Feb. 2021, [Peer-reviewed]
English, Scientific journal, The reactivity of a fuel assembly including burnable absorbers can become the largest at low fuel burnup, so the accuracy of reactivity calculations for such systems is important. To investigate this issue, the impact of gadolinium isotopes' nuclear data on neutron multiplication factor k is quantified. Sensitivities of k to nuclear data are calculated from 0 to 20 GWD/t for a BWR 3 x 3 multicell model. Sensitivity to gadolinium-157 (n, gamma) cross section becomes the largest at the zero burnup. Sensitivity to gadolinium-155 (n, gamma) cross sections takes the two largest values and the second one is observed around fuel burnup where the reactivity reaches its peak. Sensitivities are also calculated for BWR and PWR assemblies, and similar trends are observed. Finally, nuclear data-induced uncertainties of k are quantified. Gadolinium-157 contribution is the largest at zero burnup, and gadolinium-155 contribution is relatively important around fuel burnup corresponding to the reactivity peak. (C) 2020 Elsevier Ltd. All rights reserved.

Sensitivity and uncertainty analyses of fission product nuclide inventories for passive gamma spectroscopy
Go Chiba, Keisuke Honta
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 57, 12, 1265, 1275, TAYLOR & FRANCIS LTD, Dec. 2020, [Peer-reviewed]
English, Scientific journal, The passive gamma spectroscopy (PGS) is a useful technique to extract information on spent nuclear fuels without any destructive actions. This method requires a correlation between number densities (NDs) of target nuclides, and it is generally estimated by numerical simulation. Therefore, the prediction accuracy of these nuclide generations is one of the key issues in PGS. Nuclear data used in nuclear fuel depletion calculations is one of the dominant uncertainty sources, so we quantify nuclear data-induced uncertainties of NDs of six fission product nuclides, which are important in PGS: Ce-144, Cs-134, -137, Ru-106, Sb-125, and Eu-154. Generation mechanisms of these nuclides are quantitatively investigated through sensitivities of these NDs to nuclear data. With the sensitivities and covariance data of nuclear data, uncertainties of NDs of these nuclides are quantified. The uncertainties of Ce-144, Cs-137, and Ru-106 are less than 2%, and that of Sb-125 is around 6%. In these uncertainties, fission yield uncertainties are dominant. On the Cs-134 and Eu-154 generations, total uncertainties are around 5% and uncertainties of (n,) cross-sections are dominant.

A comparative study of the delta-Eddington and Galerkin quadrature methods for highly forward scattering of photons in random media
Hiroyuki Fujii, Go Chiba, Yukio Yamada, Yoko Hoshi, Kazumichi Kobayashi, Masao Watanabe
JOURNAL OF COMPUTATIONAL PHYSICS, 423, ACADEMIC PRESS INC ELSEVIER SCIENCE, Dec. 2020, [Peer-reviewed]
English, Scientific journal, A versatile and accurate treatment for the highly forward-peaked phase function in the three-dimensional (3D) radiative transfer equation (RTE) based on the discrete ordinates method (DOM) is crucial for biomedical optics. Our first objective was to compare the delta-Eddington (dE) and Galerkin quadrature (GQ) methods. The dE method decomposes the phase function into a purely forward-peaked component and the other component, and expands the other component by Legendre polynomials as well as the finite order Legendre expansion (FL) method does. The GQ method conducts the weighting procedure in addition to the Legendre expansion. Although it was reported that both methods can provide the accurate results for calculations of the RTE, the versatility of both methods is still unclear. The second objective was to examine a possibility of a conjunction of the GQ method with the dE method, called as the GQ-dE method, which has the advantages of both methods. We examined numerical errors in the moment conditions of the phase function using the FL, dE, GQ and GQ-dE methods at various types and orders of the quadrature sets, mainly in the region of the errors induced by the angular discretization using the DOM. The errors were reduced by the dE method from those by the FL method, however the error reduction depended on the types and orders of the quadrature sets. Meanwhile, the errors were significantly reduced by the GQ and GQ-dE methods, regardless of the quadrature sets. We also verified the numerical calculations of the time-dependent 3D RTE by the analytical solution of the RTE for homogeneous media in the region of the scattering length scale, where the highly forward-peaked phase function strongly influences the RTE-results. The errors in the RTE-results were similar to those in the moment conditions. Our results suggest the higher versatility and accuracy of the GQ and GQ-dE methods than those of the FL and dE methods. (C) 2020 Elsevier Inc. All rights reserved.

Quantification of integral data effectiveness using the concept of active subspace in evaluated nuclear data validation
Go Chiba, Daichi Imazato
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 57, 11, 1245, 1255, TAYLOR & FRANCIS LTD, Nov. 2020, [Peer-reviewed]
English, Scientific journal, In order to know the performance of evaluated nuclear data in reactor physics or radiation shielding calculations, its benchmark testing with integral data is mandatory. Nowadays we have a huge amount of integral data, but some of them are quite similar to each other. We need to know the independency of available integral data, and also to choose a proper set of a limited number of integral data for benchmark calculations. Furthermore, it is beneficial to know how effective a set of integral data is for independent validation of each nuclear data in performing the validation test of nuclear data. In order to quantify the effectiveness of integral data in nuclear data validation, we propose several methods based on a concept of the active subspace. With the proposed methods, we can quantify the independency of a set of integral data, choose a minimum set of proper integral data, and quantify the possibility of independent validation of nuclear data from a set of integral data. These methods are adopted to a set of fictitious integral data and a set of actual integral data including experimental data aboutand reaction rate ratio. Through these applications, effectiveness of these integral data has been successfully quantified. Furthermore, the proposed concept is utilized to interpret the nuclear data compensation effect, which has been recently discussed in the community of nuclear data.

Neutron Generation Time in Highly-Enriched Uranium Core at Kyoto University Critical Assembly
Cheol Ho Pyeon, Masao Yamanaka, Tomohiro Endo, Go Chiba, Willem F. G Van Rooijen, Kenichi Watanabe
Nuclear Science and Engineering, 194, 12, 1, 12, Informa UK Limited, 07 Jul. 2020, [Peer-reviewed]
Scientific journal

Numerical benchmark problem of solid-moderated enriched-uranium-loaded core at Kyoto university critical assembly
Go Chiba, Tomohiro Endo
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 57, 2, 187, 195, TAYLOR & FRANCIS LTD, Feb. 2020, [Peer-reviewed]
English, Scientific journal, Useful and valuable measurement data obtained at the solid-moderated core for the development of accelerator-driven systems (ADSs) have been accumulated at the Kyoto University Critical Assembly (KUCA), and some of them have been open to the public. In order to efficiently utilize these data, experimental analyses with deterministic calculation procedures are helpful. In the present manuscript, a numerical benchmark problem is established. This benchmark problem can be utilized by users of the ADS-related measurement data obtained at the KUCA A-core to verify their own numerical tools devoted to experimental analyses. Material and geometrical specifications with reference solutions obtained by a continuous-energy Monte Carlo code MVP-II are provided. In addition, numerical results obtained by a deterministic code system CBZ are also presented as an example. Through careful investigation about discretization on space and angle, guideline for proper discretization is provided. The CBZ results tend to underestimate the reference Monte Carlo solutions about 0.5% increment k/kk', and calculations of simplified core models suggest that this is caused by neutron leakage treatment in finite systems or resonance self-shielding treatment in CBZ.

Nuclear data-induced uncertainty quantification of prompt neutron decay constant based on perturbation theory for ADS experiments at KUCA
Tomohiro Endo, Kenichi Watanabe, Go Chiba, Masao Yamanaka, Willem Frederik Geert van Rooijen, Cheol Ho Pyeon
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 57, 2, 196, 204, TAYLOR & FRANCIS LTD, Feb. 2020, [Peer-reviewed]
English, Scientific journal, In experimental benchmarks of the accelerator-driven system (ADS) conducted at the Kyoto University Critical Assembly (KUCA), the prompt neutron decay constant was measured using two types of pulsed neutron sources, i.e. a D-T neutron source and a spallation neutron source driven by a 100-MeV proton beam. The measurement results of are useful information to validate the numerical results predicted by the prompt -eigenvalue calculation. In this study, the numerical analysis of using a multi-energy group S-N neutron transport code was carried out for the uranium-lead zoned experimental cores. To reduce the discretization error owing to the deterministic code, the KUCA geometry was modelled in detail as a three-dimensional heterogeneous plate-by-plate geometry, and an improved variant of EON quadrature was utilized. In addition, the sensitivity coefficients of with respect to nuclear data were efficiently evaluated by first-order perturbation theory, followed by nuclear data-induced uncertainty quantification based on the 56 neutron-energy group SCALE covariance library. Consequently, the numerical results of were validated successfully by the experimental results of the pulsed neutron source method, compared with the range of the nuclear data-induced uncertainties.

Spatially-dependent nuclear reactor kinetic calculations with the explicit fission product model
Katagiri Koji, Chiba Go
ANNALS OF NUCLEAR ENERGY, 133, 202, 208, Nov. 2019, [Peer-reviewed]

Experimental analyses of beta eff/? in accelerator-driven system at Kyoto University Critical Assembly
Yamanaka Masao, Pyeon Cheol Ho, Endo Tomohiro, Watanabe Kenichi, Chiba Go, van Rooijen Willem Frederik Geert
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 57, 2, 205, 215, Informa UK Limited, 26 Sep. 2019, [Peer-reviewed]
Scientific journal

Numerical treatment of highly forward scattering on radiative transfer using the delta-M approximation and Galerkin quadrature method
Hiroyuki Fujii, Go Chiba, Yukio Yamada, Yoko Hoshi, Kazumichi Kobayashi, Masao Watanabe
Proceedings of the 9th International Symposium on Radiative Transfer, RAD-19, SM05, 261, 268, Sep. 2019, [Peer-reviewed]
English, International conference proceedings

First nuclear transmutation of Np-237 and Am-241 by accelerator-driven system at Kyoto University Critical Assembly
Pyeon Cheol Ho, Yamanaka Masao, Oizumi Akito, Fukushima Masahiro, Chiba Go, Watanabe Kenichi, Endo Tomohiro, Van Rooijen Wilfred G, Hashimoto Kengo, Sakon Atsushi, Aizawa Naoto, Kuriyama Yasutoshi, Uesugi Tomonori, Ishi Yoshihiro
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 56, 8, 684, 689, TAYLOR & FRANCIS LTD, 03 Aug. 2019, [Peer-reviewed]
English, Scientific journal, This study demonstrates, for the first time, the principle of nuclear transmutation of minor actinide (MA) by the accelerator-driven system (ADS) through the injection of high-energy neutrons into the subcritical core at the Kyoto University Critical Assembly. The main objective of the experiments is to confirm fission reactions of neptunium-237 (Np-237) and americium-241 (Am-241), and capture reactions of Np-237. Subcritical irradiation of Np-237 and Am-241 foils is conducted in a hard spectrum core with the use of the back-to-back fission chamber that obtains simultaneously two signals from specially installed test (Np-237 or Am-241) and reference (uranium-235) foils. The first nuclear transmutation of Np-237 and Am-241 by ADS soundly implemented by combining the subcritical core and the 100 MeV proton accelerator, and the use of a lead-bismuth target, is conclusively demonstrated through the experimental results of fission and capture reaction events.

Combination of sensitivity-based and random sampling-based methodologies for efficient uncertainty quantification calculations with control variates method
Nihira Shunsuke, Chiba Go
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 19 Jun. 2019, [Peer-reviewed]

Revisiting mini-max polynomial approximation method for nuclear fuel depletion calculation　　　　　　　　　　　　　　　
Go Chiba, Yasunori Ohoka, Kento Yamamoto, Hiroaki Nagano
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019, 2401, 2410, American Nuclear Society, 2019
English, International conference proceedings, Nuclear fuel depletion calculations with a detailed nuclide transmutation chain (or a burnup chain) require advanced numerical methods and the Chebyshev rational approximation method (CRAM) has been proposed and widely used now. The mini-max polynomial approximation (MMPA) method is also a method for numerical fuel depletion calculations, and it has several advantages over CRAM. The original MMPA coefficients, which have been derived and presented in the original paper of MMPA, are determined so as to minimize approximation errors of the mini-max polynomials over a whole concerning range of a variable. In the present paper, relation between the approximation errors of the mini-max polynomials and reproduction errors of nuclide number densities after burnup is carefully investigated, and it is found that the MMPA coefficients which minimize approximation errors in a specific range of the variable can reduce the reproduction errors of nuclide number densities in comparison with the original MMPA coefficients. Through nuclear fuel burnup calculations of PWR-simulated UO2 and MOX fuel pincells and a 3×3 multicell including a burnable absorber pin, it is demonstrated that the reproduction errors of nuclide number densities after burnup can be reduced in all the cases by using the new MMPA coefficients, and reference number densities can be reproduced within 0.1% differences even with the low-order MMPA coefficients.

Accurate and efficient computation of the 3D radiative transfer equation in highly forward-peaked scattering media using a renormalization approach.
Array,Yukio Yamada, Go Chiba, Yoko Hoshi, Array, Masao Watanabe
J. Comput. Physics, 374, 591, 604, 01 Dec. 2018, [Peer-reviewed]

Uncertainty quantification works relevant to fission yields and decay data
Chiba Go, Nihira Shunsuke
EPJ NUCLEAR SCIENCES & TECHNOLOGIES, 4, 14 Nov. 2018, [Peer-reviewed]

Uncertainty quantification of criticality in solid-moderated and -reflected cores at Kyoto University Critical Assembly
Cheol Ho Pyeon, Masao Yamanaka, Makoto Ito, Go Chiba, Tomohiro Endo, Song Hyun Kim, Willem Fredrik G. van Rooijen
Journal of Nuclear Science and Technology, 55, 7, 812, 821, Taylor and Francis Ltd., 03 Jul. 2018, [Peer-reviewed]
English, Scientific journal, Uncertainty quantification is conducted for the criticality of excess reactivity and control rod worth obtained at the Kyoto University Critical Assembly (KUCA). By combining SRAC2006 and MARBLE code systems, the sensitivity coefficients of the cross sections for aluminum-27 (27Al) comprising mainly of core components are large in the solid-moderated and -reflected cores (A cores) at KUCA. Also, the uncertainty is dominant in the uranium-235 isotope by the covariance data of JENDL-4.0, and a quantitative value is about 150 pcm induced by the JENDL-4.0 data library in the KUCA A cores, whereas the covariance data of 27Al are not prepared in JENDL-4.0. Moreover, the effect of decreasing uncertainty is obtained by applying the cross-sectional adjustment method to the uncertainty analyses. From the results, a series of uncertainty quantifications is expected to clarify the uncertainty of sub-criticality in accelerator-driven system experiments with spallation neutrons in the KUCA A cores.

Research activities on nuclear reactor physics and thermal-hydraulics in Japan after Fukushima-Daiichi accident
Shuichiro Miwa, Yasunori Yamamoto, Go Chiba
Journal of Nuclear Science and Technology, 55, 6, 575, 598, Taylor and Francis Ltd., 03 Jun. 2018, [Peer-reviewed]
English, Research and development in nuclear reactor physics and thermal-hydraulics continue to be vital parts of nuclear science and technology in Japan. The Fukushima accident not only brought tremendous change in public attitudes towards nuclear engineering and technology, but also had huge influence towards the research and development culture of scientific communities in Japan. After the Fukushima accident, thorough accident reviews were completed by independent committees, namely, Tokyo Electric Power Company (TEPCO), the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and the Nuclear and Industrial Safety Agency. Reactor physics and thermal-hydraulics divisions of Atomic Energy Society of Japan (AESJ) also issued the roadmaps after the accident. As a result, lessons learned from the accident were made clear, and a number of new research activities were initiated. The present paper reviews ongoing nuclear engineering research activities in Japanese institutes, universities, and corporations, focusing on the areas in reactor physics and thermal-hydraulics since the Fukushima accident to the present date.

Experimental analysis and uncertainty quantification using random sampling technique for ADS experiments at KUCA
Tomohiro Endo, Go Chiba, Willem Frederik Geert van Rooijen, Masao Yamanaka, Cheol Ho Pyeon
Journal of Nuclear Science and Technology, 55, 4, 450, 459, Taylor and Francis Ltd., 03 Apr. 2018, [Peer-reviewed]
English, Scientific journal, Nuclear data-induced uncertainties of neutronics parameters (neutron multiplication factor keff, one-point kinetics parameters and prompt neutron decay constant α) are quantified for lead-bismuth zoned accelerator-driven system experiments at the Kyoto University Critical Assembly, in order to contribute validation for subcritical core analysis. The random sampling technique using SCALE6.2.1/Sampler/NEWT/PARTISN is utilized for the validation and the uncertainty quantification, because the random sampling technique is applicable for a problem which is not easy to apply the perturbation theory. Consequently, it is confirmed that the numerical results of α reasonably agree with the experimental ones, compared with the nuclear data-induced uncertainties. In addition, it is clarified that the nuclear data-induced correlations between α and keff and between α and neutron generation time Λ are strongly negative and positive, respectively. This fact implies that the numerical predictions of keff and Λ can be improved by the data assimilation technique using subcritical experimental results of α, which can be directly measured even for a deep subcritical system.

Perturbation theory for nuclear fuel depletion calculations with predictor–corrector method
Go Chiba
Journal of Nuclear Science and Technology, 55, 3, 290, 300, Taylor and Francis Ltd., 04 Mar. 2018, [Peer-reviewed]
English, Scientific journal, The perturbation theory for nuclear fuel depletion calculations with the predictor–corrector method is derived. This theory is implemented to a reactor physics code system CBZ, and the theory itself and its implementation are numerically verified. Sensitivities of nuclide number densities after fuel depletion with respect to nuclear data calculated with this theory are compared with reference sensitivities calculated by numerical differentiation, and good agreements are obtained. Importance of accurate angle integration on product of neutron flux and generalized adjoint neutron flux is also pointed out. Sensitivities in a 3×3 multi-cell system including a gadolinium-bearing fuel pin are calculated, and it is demonstrated that the derived theory yields accurate sensitivities even if coarse depletion time step division is adopted. The present work drastically increases the applicability of the depletion perturbation theory to actual problems.

PRESSURE DEPENDENCE OF TWO PHASE FLOW BEHAVIOR OF STAGNANT WATER IN A VERTICAL PIPE DURING STEAM INJECTION
Naoto Kitahara, Yasunori Yamamoto, Tadashi Narabayashi, Go Chiba
PROCEEDINGS OF THE 26TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2018, VOL 9, AMER SOC MECHANICAL ENGINEERS, 2018
English, International conference proceedings, Two-phase flow experiments and analysis were conducted to understand two-phase flow behavior of the water scrubbing pool of the filtered containment venting system with steam injection. In the early phase of steam injection, the water level gradually increased due to the steam condensation where the water surface was stable. After the water pool reached the saturation temperature, the diameter of bubbles increased when the bubbles moved upward in the water pool, where fluctuation of the water surface was observed. The water level increased when the scrubbing pool was pressurized by an orifice. Our simulation results showed that the decrement of the bubble velocity due to the pressurization may promoted the level swell.

Uncertainty quantification of neutron multiplication factors of light water reactor fuels during depletion
Chiba Go, Okumura Shintaro
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 55, 9, 1043, 1053, 2018, [Peer-reviewed]

Effect of instructor’s actions and attitudes on student’s motivation and discussion process in TBL class for graduate students
Shotaro Imai, Ankit Ravankar, Michiyo Shimamura, Taichi Takasuka, Go Chiba, Yasuhiro Yamanaka
International Journal of Institutional Research and Management, 1, 2, 17, 35, Sep. 2017, [Peer-reviewed]
English, Scientific journal

Analysis of the KUCA ADS Benchmarks with Diffusion Theory
W. F. G. Rooijen, T. Endo, G. Chiba, C. H. Pyeon
Progress in Nuclear Energy, 101, 243, 250, Elsevier BV, Sep. 2017, [Peer-reviewed]
English, Scientific journal

Benchmarks of subcriticality in accelerator-driven system at Kyoto University Critical Assembly
Cheol Ho Pyeon, Masao Yamanaka, Song-Hyun Kim, Thanh-Mai Vu, Tomohiro Endo, Willem Fredrik G. Van Rooijen, Go Chiba
NUCLEAR ENGINEERING AND TECHNOLOGY, 49, 6, 1234, 1239, KOREAN NUCLEAR SOC, Sep. 2017, [Peer-reviewed]
English, Scientific journal, Basic research on the accelerator-driven system is conducted by combining U-235-fueled and Th-232-loaded cores in the Kyoto University Critical Assembly with the pulsed neutron generator (14 MeV neutrons) and the proton beam accelerator (100 MeV protons with a heavy metal target). The results of experimental subcriticality are presented with a wide range of subcriticality level between near critical and 10,000 pcm, as obtained by the pulsed neutron source method, the Feynman-alpha method, and the neutron source multiplication method. (C) 2017 Korean Nuclear Society, Published by Elsevier Korea LLC.

Experimental benchmarks on kinetic parameters in accelerator-driven system with 100 MeV protons at Kyoto University Critical Assembly
Cheol Ho Pyeon, Masao Yamanaka, Tomohiro Endo, Willem Fredrik G. van Rooijen, Go Chiba
ANNALS OF NUCLEAR ENERGY, 105, 346, 354, PERGAMON-ELSEVIER SCIENCE LTD, Jul. 2017, [Peer-reviewed]
English, Scientific journal, Accelerator-driven system experiments with spallation neutrons (100 MeV protons and Pb-Bi target) are carried out in the U-235-fueled and Pb-Bi-zoned core at the Kyoto University Critical Assembly, under a subcritical state ranging between 1160 and 11,556 pcm. In these experiments, measurement of the prompt neutron decay constant and the subcriticality is conducted by the pulsed neutron source (PNS) method and the Feynman-alpha method with the use of optical fiber detectors. The experimental results successfully validate the prompt neutron decay constant and the subcriticality through the deduction of kinetic parameters by both the PNS and the a-fitting methods. The detector position dependency, neutron spectrum and subcriticality measurement methods still remain, however, in these experiments. For onward studies, the experimental benchmarks obtained from these experiments are expected to be involved in the numerical verification of subcriticality on-line monitoring, in the analysis of subcriticality uncertainty and in the deterministic approach to kinetic parameters. (C) 2017 Elsevier Ltd. All rights reserved.

Problem-based Learning and Problem Finding Among University Graduate Students
Ankit A. Ravankar, Shotaro Imai, Michiyo Shimamura, Go Chiba, Taichi Takasuka
Journal of Higher Education and Lifelong Learning, 24, 9, 20, 北海道大学高等教育推進機構, Mar. 2017, [Peer-reviewed]
English, Scientific journal, In recent years, problem-based learning (PBL) techniques have been gaining momentum inschools and university curricula around the world. The main advantage of the PBL method is that it promotescreative problem solving, improves cognition and enhances overall thought processes in learners. For mostPBL-style programmes, problem solving is at the core, although the notion of problem discovery or problemfinding is not seriously considered. In most cases, students are always presented with a structured and welldefinedproblem, but have no experience of solving an ill-structured problem or ʻwicked' problem. Thepresent study focuses on problem finding as a critical step towards developing problem solving skills inuniversity graduate students. The study aims at understanding the importance of problem formulation andcreativity, and focuses as well on our attempt to teach problem finding as an important tool in thedevelopment of creative thinking and problem solving among graduate students. The study is part of a specialgraduate programme called the Nitobe School at Hokkaido University in Japan, which started in 2015. In anactive learning classroom setting, this course is intended to support graduate students in their discovery of illstructuredproblems, help them to understand their formulation and thereby improve their problem solvingskills. We present the results of our teaching method for the first year at the Nitobe School and share ourfindings through this work.

Consistent adjustment of radioactive decay and fission yields data with measurement data of decay heat and beta-delayed neutron activities
Go Chiba
ANNALS OF NUCLEAR ENERGY, 101, 23, 30, PERGAMON-ELSEVIER SCIENCE LTD, Mar. 2017, [Peer-reviewed]
English, Scientific journal, Decay heat and delayed neutron yields, which are important physical quantities in the field of the nuclear engineering, are dependent on common nuclear data such as radioactive decay data and fission yields data of fission product nuclides. In the present study, correlations between uncertainties of these two quantities are investigated.
Nuclear data relevant to uranium-235 and plutonium-239 fissions with thermal neutron are adjusted consistently with a procedure based on Bayes' theorem using the measurement data of decay heat and delayed neutron activities. Numerical results suggest that the correlation between decay heat and delayed neutron activities uncertainties is not significant, and that independent treatments of decay heat or delayed neutron activities are possible. The effect of the consistent treatment of decay heat and delayed neutron activities is, however, observed in the adjustment results in some nuclear data such as uranium-235 thermal fission yields of yttrium-100m and zirconium-100. (C) 2016 Elsevier Ltd. All rights reserved.

Feasibility study of decay heat uncertainty reduction using nuclear data adjustment method with experimental data
Yosuke Kawamoto, Go Chiba
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 54, 2, 213, 222, TAYLOR & FRANCIS LTD, 2017, [Peer-reviewed]
English, Scientific journal, The accurate prediction of the decay heat is essential, especially for nuclear power plant safety purposes. However, it is known that the decay heat predicted by nuclear fuel burn-up calculations is uncertain because of uncertainty of nuclear data employed in the calculations. If the decay heat uncertainty can be reduced, the safety margin of the predicted decay heat can also be reduced, and feasible design ranges of various types of equipments related to the decay heat can be extended. In the present study, we use the nuclear data adjustment method for the decay heat uncertainty reduction with several types of the experimental data. As a result, we clarify that the decay heat uncertainty with short- and long-term cooling periods can be reduced by this method with appropriate experimental data.

Development of a fuel depletion sensitivity calculation module for multi-cell problems in a deterministic reactor physics code system CBZ
Go Chiba, Yosuke Kawamoto, Tadashi Narabayashi
ANNALS OF NUCLEAR ENERGY, 96, 313, 323, PERGAMON-ELSEVIER SCIENCE LTD, Oct. 2016, [Peer-reviewed]
English, Scientific journal, A new functionality of fuel depletion sensitivity calculations is developed as one module in a deterministic reactor physics code system CBZ. This is based on the generalized perturbation theory for fuel depletion problems. The theory for fuel depletion problems with a multi-layer depletion step division scheme is described in detail. Numerical techniques employed in actual implementation are also provided. Verification calculations are carried out for a 3 x 3 multi-cell problem consisting of two different types of fuel pins. It is shown that the sensitivities of nuclide number densities after fuel depletion with respect to the nuclear data calculated by the new module agree well with reference sensitivities calculated by direct numerical differentiation. To demonstrate the usefulness of the new module, fuel depletion sensitivities in different multi-cell arrangements are compared and non-negligible differences are observed. Nuclear data-induced uncertainties of nuclide number densities obtained with the calculated sensitivities are also compared. (C) 2016 Elsevier Ltd. All rights reserved.

Advanced Bondarenko method for resonance self-shielding calculations in deterministic reactor physics code system CBZ
Go Chiba, Tadashi Narabayashi
ANNALS OF NUCLEAR ENERGY, 96, 277, 286, PERGAMON-ELSEVIER SCIENCE LTD, Oct. 2016, [Peer-reviewed]
English, Scientific journal, The advanced Bondarenko method for resonance self-shielding calculations is devised and proposed. This method is based on three numerical methods; the Bell factor optimization for accurate fuel escape probability representation, extension of resonance interference factors and correction factors for current-weighted total cross sections. A 107-group library for light water reactor applications based on the advanced Bondarenko method is generated for a reactor physics code system CBZ. Performance of the CBZ code with this 107-group library is examined against a suit of light water reactor cell problems. The infinite neutron multiplication factors calculated with CBZ agree with reference continuous-energy Monte Carlo solutions within 0.15%Delta k/kk' differences, and no significant biases on fuel compositions and geometrical specifications are observed. Energy-averaged cross sections are also examined. Numerical tests reveal that significant accuracy improvements in resonance self-shielding calculations are realized by adopting the advanced Bondarenko method without any significant increase of computational burden. (C) 2016 Elsevier Ltd. All rights reserved.

Automatic construction of a simplified burn-up chain model by the singular value decomposition
Takanori Kajihara, Masashi Tsuji, Go Chiba, Yosuke Kawamoto, Yasunori Ohoka, Tadashi Ushio
ANNALS OF NUCLEAR ENERGY, 94, 742, 749, PERGAMON-ELSEVIER SCIENCE LTD, Aug. 2016, [Peer-reviewed]
English, Scientific journal, Nuclear reactor design analysis often requires a simplified, or reduced-order, burn-up chain model to reduce computation time. It is difficult to construct the reduced-order burn-up chain because it requires engineers to have highly skilled techniques and in-depth knowledge into burn-up processes. This paper develops an algorithm for automatically constructing a reduced-order burn-up chain model from a detailed model using the singular value decomposition (SVD). In our approach, we prepare a detailed burn-up chain matrix A, and an extraction matrix C, which extracts important nuclides for specific purposes such as the evaluation of neutron multiplication factor. First, the nuclides extracted by C are specified as the first candidate nuclides of the reduced-order burn-up model. Then, by applying the SVD to C, we can obtain the first information transfer matrix F-12((1)), which defines the relationship between the first candidate nuclides and remaining nuclides. In the next place, by applying SVD to F-12((1)), we can obtain additional candidate nuclides for the reduced-order burn-up chain model from the remaining nuclides. We repeat this process until the norm of the information transfer matrix is sufficiently close to zero. Finally, all candidate nuclides chosen through these simplification processes are adopted as a reduced order burn-up chain model. As a test case, we reduce a detailed burn-up chain model consisting of 1421 nuclides to a model of 204 nuclides. We can use the resulting reduced-order model to calculate the burn-up of light water reactor fuel cells with a high degree of accuracy. (C) 2016 Elsevier Ltd. All rights reserved.

DEVELOPMENT OF HIGH EFFICIENCY CONTAINMENT VENTING SYSTEM BY USING AgX
Tadashi Narabayashi, Yuuhei Sugano, Hiroki Imaeda, Go Chiba, Nobuaki Sato, Koji Endo, Toshiki Kobayashi
PROCEEDINGS OF THE 24TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2016, VOL 3, AMER SOC MECHANICAL ENGINEERS, 2016, [Peer-reviewed]
English, International conference proceedings, Fukushima Daiichi NPP accident would be terminated, if sufficient accident countermeasures, such as water proof door, mobile power, etc [1, 2]. In case of Europe, it had already installed the heat removal system and filtered containment venting system (FCVS) from the lessons of TMI and Chernobyl Accidents. The new regulatory standard in Japan, the filtered vent system (FCVS) should be installed, and prevent the radioactive material in case of the severe accident and the overpressure breakage prevention of a primary containment vessel (PCV) and also the robustization of the FCVS.
The authors examined the severe accident process in the 2nd unit of Fukushima Daiichi NPS, and found the vent by FCVS should be done before water injection into the core. The PCV spray and water injection into the pedestal basement should be also the countermeasures to the severe accident. Countermeasures for an intentional aircraft collision should be installed too.
Upon occurrence of a severe accident (SA), vent gas with radioactive fission products is blown out to a scrubbing pool through numerous venturi nozzles. Mist in steam moves upward to a metal fiber filter through a multi-hole baffle plate. After the mist is removed by that filter, radioactive methyl iodine (CH3I) is captured on the surface of a molecular sieve or AgX, made from zeolite particles with silver coating.
A FCVS visualized test facility was installed at Hokkaido University. An AgX filter is used down-stream of the scrubbing pool and metal fiver filter. Thickness of AgX filter is very important parameter to obtain enough decontamination factor (DF). The DF for the radioactive iodine exceeds 10,000 at bed depth (AgX filter thickness) greater than 75mm.

Nuclear data-induced uncertainty quantification of neutronics parameters of accelerator-driven system
Go Chiba, Cheol Ho Pyeon, Wilfred van Rooijen, Tomohiro Endo
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 53, 10, 1653, 1661, TAYLOR & FRANCIS LTD, 2016, [Peer-reviewed]
English, Scientific journal, Nuclear data-induced uncertainties of neutronics parameters of one accelerator-driven system concept designed by the Japan Atomic Energy Agency are quantified. The variance-covariance data provided in the JENDL-4.0 library are used. Uncertainties are quantified for effective neutron multiplication factor, subcritical neutron multiplication rate, a family of delayed neutron fractions, power peaking and coolant void reactivity at several operational states. Inter-cycle and inter-parameter correlation matrices and detailed information such as nuclide-wise and nuclear data-wise uncertainties are also provided.

Uncertainty quantification of total delayed neutron yields and time-dependent delayed neutron emission rates in frame of summation calculations
Go Chiba, Tadashi Narabayashi
ANNALS OF NUCLEAR ENERGY, 85, 846, 855, PERGAMON-ELSEVIER SCIENCE LTD, Nov. 2015, [Peer-reviewed]
English, Scientific journal, To contribute to further improvements of the nuclear data related to delayed neutron emissions, uncertainty quantification calculations for total delayed neutron yields, (nu) over bar (d), and time-dependent delayed neutron emission rates after a burst fission have been carried out. Those are based on the summation calculations with fundamental nuclear data taken from JENDL/FPD-2011 and a partly-modified JENDL/FPY-2011. Sensitivities required for uncertainty propagation calculations are obtained efficiently by the help of the generalized perturbation theory for time-dependent problems.
It is found that (nu) over bar (d) and neutron emission rates after a burst fission obtained in frame of summation calculations generally agree with the JENDL-4.0 evaluations within 2 sigma of nuclear data-induced uncertainty. While further improvements of the fundamental nuclear data are crucial, application of the summation calculations to actual problems is now not unrealistic, and further efforts from the application side are helpful. (C) 2015 Elsevier Ltd. All rights reserved.

Important fission product nuclides identification method for simplified burnup chain construction
Go Chiba, Masashi Tsuji, Tadashi Narabayashi, Yasunori Ohoka, Tadashi Ushio
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 52, 7-8, 953, 960, TAYLOR & FRANCIS LTD, Aug. 2015, [Peer-reviewed]
English, Scientific journal, A method of identifying important fission product (FP) nuclides which are included in a simplified burnup chain is proposed. This method utilizes adjoint nuclide number densities and contribution functions which quantify the importance of nuclide number densities to the target nuclear characteristics: number densities of specific nuclides after burnup. Numerical tests with light water reactor (LWR) fuel pin-cell problems reveal that this method successfully identifies important FP nuclides included in a simplified burnup chain, with which number densities of target nuclides after burnup are well reproduced. A simplified burnup chain consisting of 138 FP nuclides is constructed using this method, and its good performance for predictions of number densities of target nuclides and reactivity is demonstrated against LWR pin-cell problems and multi-cell problem including gadolinium-bearing fuel rod.

Numerical solution of matrix exponential in burn-up equation using mini-max polynomial approximation
Yosuke Kawamoto, Go Chiba, Masashi Tsuji, Tadashi Narabayashi
ANNALS OF NUCLEAR ENERGY, 80, 219, 224, PERGAMON-ELSEVIER SCIENCE LTD, Jun. 2015, [Peer-reviewed]
English, Scientific journal, Nuclear fuel burn-up depletion calculations are essential to compute the nuclear fuel composition transition. In the burn-up calculations, the matrix exponential method has been widely used. In the present paper, we propose a new numerical solution of the matrix exponential, a Mini-Max Polynomial Approximation (MMPA) method. This method is numerically stable for burn-up matrices with extremely short half-lived nuclides as the Chebyshev Rational Approximation Method (CRAM), and it has several advantages over CRAM. We also propose a multi-step calculation, a computational time reduction scheme of the MMPA method, which can perform simultaneously burn-up calculations with several time periods. The applicability of these methods has been theoretically and numerically proved for general burn-up matrices. The numerical verification has been performed, and it has been shown that these methods have high precision equivalent to CRAM. (C) 2015 Elsevier Ltd. All rights reserved.

Accident analysis of fukushima daiichi nuclear power station unit 1
Masahide Kobayashi, Tadashi Narabayashi, Masashi Tuji, Go Chiba, Yasunori Nagata, Tomohiro Shimoe
Transactions of the Atomic Energy Society of Japan, 14, 1, 12, 24, Atomic Energy Society of Japan, 2015, [Peer-reviewed]
Japanese, Scientific journal, As a result of the Great East Japan Earthquake that occurred on 11 March 2011, all AC and DC power at the Fukushima Daiichi NPP units 1 to 3 were lost soon after the tsunami. The core cooling function was lost, and the cores of units 1 to 3 were damaged. The purpose of this work is to clarify the progress of the accident in unit 1, which was damaged the earliest among the 3 units. Therefore, an original severe accident analysis code was developed, and the progress of the accident was evaluated from the analysis results and the actual data. As a result, the leakage path from a pressure vessel was clarified, and some lessons and knowledge were gained.

Variance Reduction Factor of Nuclear Data for Integral Neutronics Parameters
G. Chiba, M. Tsuji, T. Narabayashi
NUCLEAR DATA SHEETS, 123, 62, 67, ACADEMIC PRESS INC ELSEVIER SCIENCE, Jan. 2015, [Peer-reviewed]
English, Scientific journal, We propose a new quantity, a variance reduction factor, to identify nuclear data for which further improvements are required to reduce uncertainties of target integral neutronics parameters. Important energy ranges can be also identified with this variance reduction factor. Variance reduction factors are calculated for several integral neutronics parameters. The usefulness of the variance reduction factors is demonstrated.

Estimation of neutronics parameter sensitivity to nuclear data in random sampling-based uncertainty quantification calculations
Go Chiba, Yosuke Kawamoto, Masashi Tsuji, Tadashi Narabayashi
ANNALS OF NUCLEAR ENERGY, 75, 395, 403, PERGAMON-ELSEVIER SCIENCE LTD, Jan. 2015, [Peer-reviewed]
English, Scientific journal, We propose a method to estimate sensitivity profiles of neutronics parameters with respect to nuclear data in random sampling-based uncertainty quantification calculations. The proposed method is tested to estimate sensitivity profiles of fast neutron systems criticalities. A high order effect in sensitivity profile estimation is found to be quite important, so a reverse sampling method is developed to mitigate the high order effect. With this reverse sampling method, detail energy group-wise sensitivity profiles can be estimated even though some fluctuations are observed in specific sensitivity profiles. Energy-integrated sensitivity profiles can be accurately calculated with the proposed method.
With the estimated sensitivity profiles, partial uncertainties, that are neutronics parameters uncertainties induced by specific nuclear data uncertainties, are also calculated. Numerical tests reveal that the proposed method reproduces quite well the reference partial uncertainties. A simple and practical partial uncertainty estimation method, which only requires a covariance matrix between neutronics parameter and nuclear data, is also tested and assessed. (C) 2014 Elsevier Ltd. All rights reserved.

RESONANCE SELF-SHIELDING EFFECT IN UNCERTAINTY QUANTIFICATION OF FISSION REACTOR NEUTRONICS PARAMETERS
Go Chiba, Masashi Tsuji, Tadashi Narabayashi
NUCLEAR ENGINEERING AND TECHNOLOGY, 46, 3, 281, 290, KOREAN NUCLEAR SOC, Jun. 2014, [Peer-reviewed]
English, Scientific journal, In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

Photon transport effect on intra-subassembly thermal power distribution in fast reactor
Go Chiba, Masashi Tsuji, Tadashi Narabayashi
ANNALS OF NUCLEAR ENERGY, 65, 41, 46, PERGAMON-ELSEVIER SCIENCE LTD, Mar. 2014
English, Scientific journal, In order to accurately predict intra-subassembly thermal power distribution in a fast reactor, neutron and photon transport calculations are carried out with a multi-purpose reactor physics calculation code system CBZ. All the fission fragment nuclide are treated explicitly during fuel depletion, and irradiation time-dependent energy spectra of delayed fission gamma-rays emitted from all the fission fragment nuclides are precisely simulated. Time-dependent delayed beta-ray emission and transmutations of fission fragment nuclide by neutron-nuclide reactions are also taken into account. A fuel subassembly model of Japanese prototype fast reactor Monju is used for numerical calculations, and their two-dimensional geometric feature is precisely modeled by a ray-tracing-based collision probability method implemented in CBZ. When the photon transport is considered, total thermal powers in fissile material regions are reduced by about 1.5% except at the beginning of fuel depletion. (C) 2013 Elsevier Ltd. All rights reserved.

Nuclear data sensitivity analysis for isotopic generation using jendl-4.0, endf/b-vii.1 and jeff-3.1.1　　　　　　　　　　　　　　　
Yosuke Kawamoto, Go Chiba, Masashi Tsuji, Tadashi Narabayashi
Proceedings of the International Conference on Physics of Reactors, PHYSOR 2014, Japan Atomic Energy Agency, JAEA, 2014
English, International conference proceedings, In burn-up calculations, accuracy of isotopic generation prediction strongly depends on the nuclear data, such as neutron cross sections, fission yields and decay constants. In this study, we perform burn-up calculations using JENDL-4.0, ENDF/B-VII.1 and JEFF-3.1.1, and compare the calculation results with PIE (Post Irradiation Examination) data. We also perform burn-up sensitivity analysis based on the generalized perturbation theory to clarify the cause of differ-ence on isotopic generation between the libraries. As a result, there are large discrepancies between JEFF-3.1.1 and the others generally. Furthermore, we clarify their causes for each nuclide and energy group. For neutron cross sections, some nuclides have large discrepancies between JEFF-3.1.1 and the others, and they give large impacts on specific isotopic genera-tion predictions. On fission yields, ones from Pu-239 and Pu-241 have large discrepancies between JEFF-3.1.1 and the others, and they give large impacts on specific isotopic generation predictions, especially Gd-160. Decay constant discrepancies do not give any large impacts on isotopic generation predictions.

Recent activities in the field of reactor physics
Masahiro Tatsumi, Go Chiba
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 51, 9, 1161, 1163, TAYLOR & FRANCIS LTD, 2014, [Peer-reviewed]
English, As the basis and fundamentals of nuclear technology, reactor physics has an important role to play; recent requirements for reliability and accountability to realize a higher level of safety have been encouraging researchers and engineers to study and develop more advanced and sophisticated numerical methods and calculation codes. Many of the outstanding research and developments are presented in scientific journals, including the Journal of Nuclear Science and Technology. Some topics have been covered in this summary from the latest activities in the field of reactor physics.

Sensitivity and uncertainty analysis for reactor stable period induced by positive reactivity using one-point adjoint kinetics equation
Go Chiba, Masashi Tsuji, Tadashi Narabayashi
Journal of Nuclear Science and Technology, 50, 12, 1150, 1160, 01 Dec. 2013
English, Scientific journal, In order to better predict the kinetic behavior of a nuclear fission reactor, improvement of delayed neutron parameters is essential. Since it is required to establish a path from the microscopic nuclear data to the macroscopic delayed neutron parameters for the improvement, the present paper identifies important nuclear data for reactor kinetics. Sensitivities of the reactor stable period, which describes reactor kinetic behavior, to microscopic nuclear data such as independent fission yields, decay constants and decay branching ratios are calculated efficiently by using the adjoint kinetics equation. Furthermore, nuclide-wise and nuclear data-wise uncertainties of the reactor stable period are quantified using the variance data given in the nuclear data file, and the nuclear data that require further improvement are identified.The results obtained through the present study are quite helpful, and can be a driving force for further nuclear physics studies. © 2013 Atomic Energy Society of Japan. All rights reserved.

Uncertainty quantification of neutronic parameters of light water reactor fuel cells with JENDL-4.0 covariance data
Go Chiba, Masashi Tsuji, Tadashi Narabayashi
Journal of Nuclear Science and Technology, 50, 7, 751, 760, 01 Jul. 2013
English, Scientific journal, Neutronic parameter uncertainty induced by nuclear data uncertainty is quantified for several light water reactor fuel cells composed of different combinations of fissile/fertile nuclides. The covariance data given in JENDL-4.0 are used as the nuclear data uncertainty, and uncertainty propagation calculations are carried out using sensitivity coefficients calculated with the generalized perturbation theory for burnup-related neutronic parameters. It is found that main contributors of nuclear data uncertainty to the neutronic parameter uncertainty are the uranium-238 capture cross section in a uranium-oxide fuel cell, and the plutonium-240 and plutonium-241 capture cross sections and fission spectrum of fissile plutonium isotopes in a uranium-plutonium mixed-oxide fuel cell. It is also found that thorium-232 capture cross section uncertainty is a dominant source of neutronic parameter uncertainty in thorium-uranium and thorium-plutonium mixed-oxide fuel cells. It should be emphasized that precise and detail information of component-wise uncertainties can be obtained by virtue of the adjoint-based sensitivity calculation methodology. Furthermore, cross-correlations are evaluated for each fuel cell, and strong correlations among the same parameters at the beginning of cycle and at the end of cycle and among different parameters are observed. © 2013 Taylor and Francis Group, LLC.

Important fission product nuclides identification method for simplified burnup-chain construction　　　　　　　　　　　　　　　
Go Chiba, Masashi Tsuji, Tadashi Narabayashi, Yasunori Ohoka, Tadashi Ushio
Transactions of the American Nuclear Society, 109, 2, 1299, 1300, American Nuclear Society, 2013
English, International conference proceedings

Efficient fission neutron spectrum matrix representation by singular value decomposition technique
Go Chiba, Akio Yamamoto, Masashi Tsuji, Tadashi Narabayashi
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 49, 7, 748, 753, TAYLOR & FRANCIS LTD, Jul. 2012
English, Scientific journal, In order to treat efficiently a huge fission neutron spectrum (FNS) matrix in a criticality calculation, the singular value decomposition (SVD) technique is introduced to an FNS matrix representation. The required number of SVD components for reconstruction of an FNS matrix is expected to be small since an incident neutron energy dependence of FNS is not so significant. The proposed technique of an SVD-based representation for a fission source term is tested in several fast critical systems. Through an observation of critical eigenvalue dependence on the number of considered SVD components, only six or seven components are required to obtain a critical eigenvalue which agrees with the reference solution within 10(-4) dk/kk'. It is also confirmed that a small reactivity effect caused by neutron spectrum shifting can be accurately calculated with the proposed technique.

A note on application of superhomogeneisation factors to integro-differential neutron transport equations
Go Chiba, Masashi Tsuji, Ken-ichiro Sugiyama, Tadashi Narabayashi
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 49, 1-2, 272, 280, TAYLOR & FRANCIS LTD, Jan. 2012
English, Scientific journal, The present article focuses on the application of the SPH factor method to the integro-differential neutron transport equation. While leakage-related parameters are arbitrarily corrected by the SPH factors, the correction procedure for these parameters affects the calculation accuracy. We treat two correction procedures named the simultaneous correction and the direct correction, and compare them with each other in one-dimensional colorset assembly problems. Through numerical testing, we find that the simultaneous SPH correction gives better accuracy than the direct SPH correction, and the higher-order SPH-corrected calculations show better accuracy than the low-order ones. Furthermore, to consider the flux discontinuity between different types of assemblies, the improved SPH method proposed by Yamamoto and the SPH method with the Selengut normalization condition are also tested. Numerical results reveal that the both methods significantly improve the calculation accuracy and that the latter method is more robust than the former method.

JENDL-4.0 Benchmarking for Effective Delayed Neutron Fraction of Fast Neutron Systems
Go Chiba, Masashi Tsuji, Ken-ichiro Sugiyama, Tadashi Narabayashi
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 48, 12, 1471, 1477, TAYLOR & FRANCIS LTD, Dec. 2011
English, Scientific journal, The performance of the latest Japanese evaluated nuclear data library JENDL-4.0 for the prediction of effective delayed neutron fraction beta(eff) is assessed using experimental data of a wide range of fast neutron systems. Covariance data of JENDL-4.0 are used to quantify nuclear-data-induced uncertainties. Calculations with other libraries, JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1, are also carried out for a quantitative comparison. JENDL-4.0 results in good agreement between calculation and experimental values within total uncertainties, and consistency between the differential nuclear data and integral experimental data is confirmed. While the other libraries also show good performance for beta(eff) prediction, there are small differences in the predicted values of beta(eff) among different libraries and ENDF/B-VII.0 gives the most stable results. Furthermore, a simple and convenient procedure to calculate sensitivity profiles of beta(eff) to nuclear data is proposed.

JENDL-4.0 Benchmarking for Effective Delayed Neutron Fraction of Fast Neutron Systems
Go Chiba, Masashi Tsuji, Ken-ichiro Sugiyama, Tadashi Narabayashi
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 48, 12, 1471, 1477, TAYLOR & FRANCIS LTD, Dec. 2011, [Peer-reviewed]
English, Scientific journal, The performance of the latest Japanese evaluated nuclear data library JENDL-4.0 for the prediction of effective delayed neutron fraction beta(eff) is assessed using experimental data of a wide range of fast neutron systems. Covariance data of JENDL-4.0 are used to quantify nuclear-data-induced uncertainties. Calculations with other libraries, JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1, are also carried out for a quantitative comparison. JENDL-4.0 results in good agreement between calculation and experimental values within total uncertainties, and consistency between the differential nuclear data and integral experimental data is confirmed. While the other libraries also show good performance for beta(eff) prediction, there are small differences in the predicted values of beta(eff) among different libraries and ENDF/B-VII.0 gives the most stable results. Furthermore, a simple and convenient procedure to calculate sensitivity profiles of beta(eff) to nuclear data is proposed.

Benchmark Calculations of Sodium-Void Experiments with Uranium Fuels at the Fast Critical Assembly FCA
FUKUSHIMA MASAHIRO, KITAMURA YASUNORI, KUGO TERUHIKO, YAMANE TSUYOSHI, ANDOH MASAKI, CHIBA GO, ISHIKAWA MAKOTO, OKAJIMA SHIGEAKI
Prog Nucl Sci Technol (Web), 2, WEB ONLY 306-311, Oct. 2011
English

On Effective Delayed Neutron Fraction Calculations with Iterated Fission Probability
Go Chiba, Yasunobu Nagaya, Takamasa Mori
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 48, 8, 1163, 1169, TAYLOR & FRANCIS LTD, Aug. 2011
English, Scientific journal, The iterated fission probability (IFP) is a quantity proportional to the asymptotic power level originated by a neutron introduced to a reactor. The effective delayed neutron fraction beta(eff) can be accurately calculated by the continuous-energy Monte Carlo method using IFP if a sufficiently large number of generations is considered to obtain the asymptotic state. In order to deterministically quantify the required number of generations in the IFP-based beta(eff) calculations, the concept of the generation-dependent importance functions is introduced to beta(eff) calculations. Furthermore, the most appropriate reactor property used in the IFP calculations, which reduces the required number of generations, is theoretically derived. Through numerical calculations, it is shown that several generations are required in the IFP-based beta(eff) calculations and that the use of the appropriate reactor property can reduce the required number of generations. An efficient procedure for the IFP-based beta(eff) calculations by the Monte Carlo method is also proposed.

JENDL-4.0 Integral Testing for Fission Systems
Keisuke Okumura, Kazuteru Sugino, Go Chiba, Yasunobu Nagaya, Kenji Yokoyama, Teruhiko Kugo, Makoto Ishikawa, Shigeaki Okajima
JOURNAL OF THE KOREAN PHYSICAL SOCIETY, 59, 2, 1135, 1140, KOREAN PHYSICAL SOC, Aug. 2011, [Peer-reviewed]
English, Scientific journal, Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of the integral data. Benchmark calculations are performed with the continuous-energy Monte Carlo code with a large number of neutron histories or with the deterministic procedure which has been developed for fast reactor analyses in japan. In the present paper, representative benehmark results are shown as a rapid report. They are the results for criticality of low enriched UO(2) or MOX fueled light water moderated systems, of uranium or plutonium fuelled solution systems, of various fast reactors, and results of PIE analyses for a PWR, spent fuel and actinoide samples irradiated in fast reactors.

Development of a Unified Cross-section Set ADJ2010 Based on Adjustment Technique for Fast Reactor Core Design
K. Sugino, M. Ishikawa, K. Yokoyama, Y. Nagaya, G. Chiba, T. Hazama, T. Kugo, K. Numata, T. Iwai, T. Jin
JOURNAL OF THE KOREAN PHYSICAL SOCIETY, 59, 2, 1357, 1360, KOREAN PHYSICAL SOC, Aug. 2011, [Peer-reviewed]
English, Scientific journal, In order to improve the prediction accuracy of core performances in the fast reactor core design study, the unified cross-section set has been developed in Japan. The unified cross-section set, which combines a wide range of integral experimental information with differential nuclear data, is produced by using the cross-section adjustment technique based on the Bayesian parameter-estimation theory. A new set ADJ2010 is currently under development. The present paper describes the results of the cross-section adjustment for ADJ2010 which is based on the JENDL-4.0 data. The evaluation of the core design accuracy for a commercial power fast reactor core is also discussed. ADJ2010 will be released soon and will be expected to be utilized for core design of future fast reactors.

On Effective Delayed Neutron Fraction Calculations with Iterated Fission Probability
Go Chiba, Yasunobu Nagaya, Takamasa Mori
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 48, 8, 1163, 1169, TAYLOR & FRANCIS LTD, Aug. 2011, [Peer-reviewed]
English, Scientific journal, The iterated fission probability (IFP) is a quantity proportional to the asymptotic power level originated by a neutron introduced to a reactor. The effective delayed neutron fraction beta(eff) can be accurately calculated by the continuous-energy Monte Carlo method using IFP if a sufficiently large number of generations is considered to obtain the asymptotic state. In order to deterministically quantify the required number of generations in the IFP-based beta(eff) calculations, the concept of the generation-dependent importance functions is introduced to beta(eff) calculations. Furthermore, the most appropriate reactor property used in the IFP calculations, which reduces the required number of generations, is theoretically derived. Through numerical calculations, it is shown that several generations are required in the IFP-based beta(eff) calculations and that the use of the appropriate reactor property can reduce the required number of generations. An efficient procedure for the IFP-based beta(eff) calculations by the Monte Carlo method is also proposed.

Application of the hierarchical domain decomposition boundary element method to the simplified P-3 equation
Go Chiba
ANNALS OF NUCLEAR ENERGY, 38, 5, 1033, 1038, PERGAMON-ELSEVIER SCIENCE LTD, May 2011, [Peer-reviewed]
English, Scientific journal, In this paper, the hierarchical domain decomposition boundary element method (HDD-BEM), which has been developed to solve the diffusion equation, is applied to the simplified P-3 (SP3) equation. The HDD-BEM solution for the SP3 equation is provided in the present paper. A computer program, ABEMIE, based on the HDD-BEM is developed, and a two-dimensional one-group anisotropic-scattering benchmark problem is solved with it to verify the present HDD-BEM for the SP3 equation.
Through numerical benchmarking, it is shown that the present method results in good agreement with the solution obtained using the existing SPN solver based on the finite element method for both eigen-value and neutron flux distribution. This benchmark result suggests that the HDD-BEM is suitable for application to the SPN equation. (C) 2011 Elsevier Ltd. All rights reserved.

JENDL-4.0 Benchmarking for Fission Reactor Applications
Go Chiba, Keisuke Okumura, Kazuteru Sugino, Yasunobu Nagaya, Kenji Yokoyama, Teruhiko Kugo, Makoto Ishikawa, Shigeaki Okajima
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 48, 2, 172, 187, TAYLOR & FRANCIS LTD, Feb. 2011, [Peer-reviewed]
English, Scientific journal, Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0.

Improvement of Tone's Method with Two-Term Rational Approximation
Akio Yamamoto, Tomohiro Endo, Go Chiba
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 48, 2, 263, 271, TAYLOR & FRANCIS LTD, Feb. 2011, [Peer-reviewed]
English, Scientific journal, An improvement of Tone's method, which is a resonance calculation method based on the equivalence theory, is proposed. In order to increase calculation accuracy, the two-term rational approximation is incorporated for the representation of neutron flux. Furthermore, some theoretical aspects of Tone's method, i.e., its inherent approximation and choice of adequate multigroup cross section for collision probability estimation, are also discussed. The validity of improved Tone's method is confirmed through a verification calculation in an irregular lattice geometry, which represents part of an LWR fuel assembly. The calculation result clarifies the validity of the present method.

Diffusion coefficients for LMFBR cells calculated with MOC and Monte Carlo methods
W. F. G. van Rooijen, G. Chiba
ANNALS OF NUCLEAR ENERGY, 38, 1, 133, 144, PERGAMON-ELSEVIER SCIENCE LTD, Jan. 2011, [Peer-reviewed]
English, Scientific journal, The present work discusses the calculation of the diffusion coefficient of a lattice of hexagonal cells with both,odium present and sodium absent conditions Calculations are performed in the framework of Lance theory (also known as fundamental mode approximation) Unlike the classical approaches our heterogeneous leakage model allows the calculation of diffusion coefficients under all conditions even if planar voids are present in the lattice Equations resulting from this model are solved using the method of characteristics (MOC) Independent confirmation of the MOC result comes from Monte Carlo calculations in which the diffusion coefficient is obtained without any of the assumptions of lattice theory It is shown by comparison to the Monte Carlo results that the MOC solution yields correct values of the diffusion coefficient under all conditions even in cases where the classic calculation of the diffusion coefficient fails This work is a first step in the development of a robust method to calculate the diffusion coefficient of lattice cells Adoption into production codes will require more development and validation of the method (C) 2010 Elsevier Ltd All rights reserved

Comparison of Monte Carlo calculation methods for effective delayed neutron fraction
Yasunobu Nagaya, Go Chiba, Takamasa Mori, Dwi Irwanto, Ken Nakajima
ANNALS OF NUCLEAR ENERGY, 37, 10, 1308, 1315, PERGAMON-ELSEVIER SCIENCE LTD, Oct. 2010, [Peer-reviewed]
English, Scientific journal, Monte Carlo calculation methods to estimate the effective delayed neutron fraction beta(eff) are investigated: one is proposed by Meulekamp et al. and the other is by Nauchi et al. It is revealed that both the methods calculate the delayed neutron fraction weighted with the importance functions defined by Kobayashi. The accuracy of the methods are also examined for several simple benchmark systems. Consequently, it is found that Meulekamp's method causes similar to 5% discrepancies in the flea values for fast systems; Nauchi's method gives good results for fast bare systems but similar to 10% discrepancies for fast reflected systems. Both the methods calculate the beta(eff) values approximately within the accuracy of similar to 2% for thermal systems. (C) 2010 Elsevier Ltd. All rights reserved.

Sensitivity Analysis of Fission Product Concentrations for Light Water Reactor Burned Fuel
Go Chiba, Keisuke Okumura, Akito Oizumi, Masaki Saito
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 47, 7, 652, 660, TAYLOR & FRANCIS LTD, Jul. 2010
English, Scientific journal, The accurate prediction of fission product concentrations (FPCs) is necessary for application of the burnup credit to nuclear facilities. In order to specify important nuclear data for the accurate prediction of FPC, we extensively evaluate the sensitivities of FPC to nuclear data with the depletion perturbation theory. The target fission products are twelve important ones for the burnup credit, Mo-95, Tc-99, Rh-103, Nd-143, Nd-145, Sm-147, Sm-149, Sm-150, Sm-152, Cs-133, Eu-153, and Gd-155. The present study successfully specifies the important nuclear data both in a UO2 cell and in a MOX cell. While the obtained sensitivities are mostly similar to each other between the UO2 and MOX cells, large differences are observed in some cases, such as the Gd-155 concentration. It is clearly shown that such differences between the UO2 and MOX cells come from differences in cumulative fission yields between U-235 and Pu-239 and differences in neutron flux energy spectra.

Sensitivity Analysis of Fission Product Concentrations for Light Water Reactor Burned Fuel
Go Chiba, Keisuke Okumura, Akito Oizumi, Masaki Saito
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 47, 7, 652, 660, TAYLOR & FRANCIS LTD, Jul. 2010, [Peer-reviewed]
English, Scientific journal, The accurate prediction of fission product concentrations (FPCs) is necessary for application of the burnup credit to nuclear facilities. In order to specify important nuclear data for the accurate prediction of FPC, we extensively evaluate the sensitivities of FPC to nuclear data with the depletion perturbation theory. The target fission products are twelve important ones for the burnup credit, Mo-95, Tc-99, Rh-103, Nd-143, Nd-145, Sm-147, Sm-149, Sm-150, Sm-152, Cs-133, Eu-153, and Gd-155. The present study successfully specifies the important nuclear data both in a UO2 cell and in a MOX cell. While the obtained sensitivities are mostly similar to each other between the UO2 and MOX cells, large differences are observed in some cases, such as the Gd-155 concentration. It is clearly shown that such differences between the UO2 and MOX cells come from differences in cumulative fission yields between U-235 and Pu-239 and differences in neutron flux energy spectra.

Neutronic calculations for steel-reflected fast critical systems with the sub-group Sn method　　　　　　　　　　　　　　　
Go Chiba, Teruhiko Kugo
International Conference on the Physics of Reactors 2010, PHYSOR 2010, 2, 1064, 1074, American Nuclear Society, 2010
English, International conference proceedings, In the present paper, we perform neutronic calculations for steel-reflected fast critical systems with the sub-group SN method. In order to extend the applicability of the sub-group SN method, we consider sub-group to sub-group transfer probabilities for in-group scattering sources. In addition, sub-group dependence of out-group scattering sources, which has been ignored in previous studies, is also taken into account. The present sub-group SN method is applied to neutronic calculations for several steel-reflected fast systems included in the ICSBEP handbook. It is shown that the present sub-group SN method reproduces quite well the reference Monte-Carlo solutions for effective multiplication factors and neutron flux spatial distributions above 0.1 MeV in reflector regions. This method, however, shows poor accuracy in neutron flux calculations for specific energy groups in which large and wide resonances exist or in which the contribution of the inelastic scattering source is large. These are the limitations of the present sub-group SN method, and remedy for them is necessary if accurate neutron flux calculations are required for such energy groups.

Evaluation of Neutron Nuclear Data on Arsenic-75 for JENDL-4
Keiichi Shibata, Go Chiba, Akira Ichihara, Satoshi Kunieda
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 47, 1, 40, 46, TAYLOR & FRANCIS LTD, Jan. 2010
English, Scientific journal, Neutron nuclear data on As-75 have been evaluated for the evaluated nuclear data library JENDL-4 in the energy region front 10(-5) eV to 20 MeV. The thermal capture cross section was updated by considering recent measurements. The statistical model was applied to calculate the cross sections above the resolved resonance region. In the calculation, coupled-channel optical model parameters were used for neutrons. Pre-equilibrium and direct-reaction processes were taken into account in addition to the compound process. The present calculations are almost consistent with available experimental data. The measured leakage neutron spectrum is well reproduced by the presently evaluated data at 14 MeV.

Evaluation of Neutron Nuclear Data on Arsenic-75 for JENDL-4
Keiichi Shibata, Go Chiba, Akira Ichihara, Satoshi Kunieda
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 47, 1, 40, 46, TAYLOR & FRANCIS LTD, Jan. 2010, [Peer-reviewed]
English, Scientific journal, Neutron nuclear data on As-75 have been evaluated for the evaluated nuclear data library JENDL-4 in the energy region front 10(-5) eV to 20 MeV. The thermal capture cross section was updated by considering recent measurements. The statistical model was applied to calculate the cross sections above the resolved resonance region. In the calculation, coupled-channel optical model parameters were used for neutrons. Pre-equilibrium and direct-reaction processes were taken into account in addition to the compound process. The present calculations are almost consistent with available experimental data. The measured leakage neutron spectrum is well reproduced by the presently evaluated data at 14 MeV.

Impact of Incident Energy Dependence of Prompt Fission Neutron Spectra on Uncertainty Analyses
Go Chiba, Yasunobu Nagaya
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 46, 10, 1000, 1003, TAYLOR & FRANCIS LTD, Oct. 2009
English, Scientific journal, This paper investigates the impact of the incident energy dependence of prompt fission neutron spectra (PFNS) on uncertainty propagation calculations. Uncertainty propagation from incident energy-dependent PFNS to criticality is formulated and its impact is evaluated numerically. It is found that the conventional procedure, in which representative PFNS covariance data for a specific incident energy are used, results in a larger PFNS-induced uncertainty than the straightforward procedure, in which different PFNS covariance data are used for each incident energy range given in the nuclear data libraries. The present study suggests that the correlation between different incident energies of PFNS has a large impact on uncertainty propagation calculation results for nuclear characteristics.

Impact of Incident Energy Dependence of Prompt Fission Neutron Spectra on Uncertainty Analyses
Go Chiba, Yasunobu Nagaya
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 46, 10, 1000, 1003, TAYLOR & FRANCIS LTD, Oct. 2009, [Peer-reviewed]
English, Scientific journal, This paper investigates the impact of the incident energy dependence of prompt fission neutron spectra (PFNS) on uncertainty propagation calculations. Uncertainty propagation from incident energy-dependent PFNS to criticality is formulated and its impact is evaluated numerically. It is found that the conventional procedure, in which representative PFNS covariance data for a specific incident energy are used, results in a larger PFNS-induced uncertainty than the straightforward procedure, in which different PFNS covariance data are used for each incident energy range given in the nuclear data libraries. The present study suggests that the correlation between different incident energies of PFNS has a large impact on uncertainty propagation calculation results for nuclear characteristics.

JENDL Actinoid File 2008
Osamu Iwamoto, Tsuneo Nakagawa, Naohiko Otuka, Satoshi Chiba, Keisuke Okumura, Go Chiba, Takaaki Ohsawa, Kazuyoshi Furutaka
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 46, 5, 510, 528, TAYLOR & FRANCIS LTD, May 2009
English, Scientific journal, JENDL Actinoid File 2008 (JENDL/AC-2008) was released in March 2008. It includes nuclear data for neutron-induced reactions for 79 nuclides from Ac (Z = 89) to Fm (Z = 100). The neutron energy range is 10(-5) eV to 20 MeV. Almost all data for 62 actinoids in JENDL-33 were revised. New evaluations were performed for 17 nuclides, which have half-lives longer than one day. A new comprehensive theoretical model code CCONE was widely used for the evaluation of cross sections and neutron emission spectra. Thermal cross sections for many nuclides were revised based on experimental data. Resonance parameters were readjusted to reproduce them. Simultaneous evaluations of fission cross sections were performed for six important nuclei. The least-squares fitting code GMA was used for the evaluation of fission cross sections for minor actinoids. In this paper, we present the evaluation methods and results of the JENDL/AC-2008.

Evaluation of sensitivity coefficients of effective multiplication factor with respect to prompt fission neutron spectrum
Yasunobu Nagaya, Ivan Kodeli, Go Chiba, Makoto Ishikawa
NUCLEAR INSTRUMENTS & METHODS IN PHYSICS RESEARCH SECTION A-ACCELERATORS SPECTROMETERS DETECTORS AND ASSOCIATED EQUIPMENT, 603, 3, 485, 490, ELSEVIER SCIENCE BV, May 2009, [Peer-reviewed]
English, Scientific journal, Sensitivity coefficients with respect to the fission neutron spectrum can be formulated in two ways: in terms of the unconstrained and constrained sensitivity coefficients. Differences in results obtained using the two sensitivity methods were investigated for the case of the perturbation of the fission spectrum, and for the uncertainty of the effective multiplication factor (k(eff)) with respect to the fission spectrum. It is shown analytically that both sensitivity coefficient methods ideally result in the same uncertainty predictions. However, if the unconstrained sensitivity coefficients are used, the zero-sum constraint for the absolute covariance matrix must be satisfied with high numerical accuracy. (C) 2009 Elsevier B.V. All rights reserved.

Calculation of Effective Delayed Neutron Fraction Using a Modified k-Ratio Method
Go Chiba
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 46, 5, 399, 402, TAYLOR & FRANCIS LTD, May 2009, [Peer-reviewed]
English, Scientific journal

JENDL Actinoid File 2008
Osamu Iwamoto, Tsuneo Nakagawa, Naohiko Otuka, Satoshi Chiba, Keisuke Okumura, Go Chiba, Takaaki Ohsawa, Kazuyoshi Furutaka
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 46, 5, 510, 528, TAYLOR & FRANCIS LTD, May 2009, [Peer-reviewed]
English, Scientific journal, JENDL Actinoid File 2008 (JENDL/AC-2008) was released in March 2008. It includes nuclear data for neutron-induced reactions for 79 nuclides from Ac (Z = 89) to Fm (Z = 100). The neutron energy range is 10(-5) eV to 20 MeV. Almost all data for 62 actinoids in JENDL-33 were revised. New evaluations were performed for 17 nuclides, which have half-lives longer than one day. A new comprehensive theoretical model code CCONE was widely used for the evaluation of cross sections and neutron emission spectra. Thermal cross sections for many nuclides were revised based on experimental data. Resonance parameters were readjusted to reproduce them. Simultaneous evaluations of fission cross sections were performed for six important nuclei. The least-squares fitting code GMA was used for the evaluation of fission cross sections for minor actinoids. In this paper, we present the evaluation methods and results of the JENDL/AC-2008.

Calculation of effective delayed neutron fraction using a modified k-ratio method
Go Chiba
Journal of Nuclear Science and Technology, 46, 5, 399, 402, 2009
English, Scientific journal, A modified k-ratio method, which is applicable to continuous-energy Monte Carlo simulations, is proposed to estimate rigorous values of the effective delayed neutron fraction. The adjoint neutron flux is used as the weight function and the the neutron transport equation at a fictitious state where the delayed neutrons are ignored is considered. The results show that the proposed k-ratio method with scaling factors close to zero results in good agreement with the references, and use of the large scaling factors make the errors larger. The errors in neutron fraction are found to be less than 0.2% for the nine fast systems. The proposed method with positive scaling factors is found to yield more accurate results than that with negative scaling factors. The proposed k-ratio method yields more accurate results for the bare systems than for the reflected systems.

JENDL actinoid file 2008 and plan of covariance evaluation
Iwamoto, Osamu, Nakagawa, Tsuneo, Otsuka, Naohiko, Chiba, Satoshi, Okumura, Keisuke, Chiba, Go
Nuclear Data Sheets, 109, 12, 2885, 2889, Dec. 2008, [Peer-reviewed]
English, JENDL Actinoid File 2008 (JENDL/AC-2008) was released in March 2008. It provides nuclear data of neutron induced nuclear reactions for actinoid nuclides from Ac to Fm. The data for 62 nuclides in JENDL-3.3 were revised and newly evaluated data for 17 nuclides, which have a half-life longer than 1 day, were added. Nuclear reaction model code CCONE was widely used for the evaluations of cross sections and energy-angular distributions of secondary neutrons in fast energy region. Covariance data for the fission and capture cross sections and the number of neutrons per fission will be evaluated for important nuclides in the JENDL/AC-2008. The evaluation methods and the results will be presented and the plan of the covariance evaluations also will be mentioned.

Development of JENDL actinoid file　　　　　　　　　　　　　　　
Iwamoto, Osamu, Nakagawa, Tsuneo, Otsuka, Naohiko, Chiba, Satoshi, Okumura, Keisuke, Chiba, Go
Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8, Sep. 2008
English, Nuclear data for neutron induced reactions with actinides from Ac to Fm have been evaluated for JENDL Actinoid File (JENDL/AC). Almost all data in JENDL-3.3 have been updated based on available experimental data and using the newly developed theoretical model code CCONE. Integral benchmark tests for fission reactors are in progress using preliminary versions of JENDL/AC. The JENDL/AC will be released in 2008.

Covariance analyses of self-shielding factor and its temperature gradient for uranium-238 neutron capture reaction
Naohiko Otuka, Atsushi Zukeran, Hideki Takano, Go Chiba, Makoto Ishikawa
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 45, 3, 195, 210, ATOMIC ENERGY SOC JAPAN, Mar. 2008, [Peer-reviewed]
English, Scientific journal, Covariances of the self-shielding factor and its temperature gradient for the uranium-238 neutron capture reaction have been evaluated from the resonance parameter covariance matrix and the sensitivity of the self-shielding factor and its temperature gradient to the resonance parameters. The resonance parameters and their covariance matrix for uranium-238 were taken from JENDL-3.3, while the sensitivity coefficients were calculated by varying resonance parameters and temperature. A set of computer code modules has been developed for the calculation of the sensitivity coefficients at numerous resonance levels. The present result shows that the correlation among resonance parameters yields a substantial contribution to the standard deviations of the self-shielding factor and its temperature gradient. In addition to the standard deviations of these quantities, their correlation matrices in the JFS-3 70 group structure are also obtained.

Sodium void reactivity worth calculations for fast critical assemblies without whole-lattice homogenization
Go Chiba, Yoichiro Shimazu
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 44, 12, 1526, 1534, ATOMIC ENERGY SOC JAPAN, Dec. 2007, [Peer-reviewed]
English, Scientific journal, In the present paper, we calculate the sodium void reactivity worth of fast critical assemblies without whole-lattice homogenization in order to reduce errors associated with lattice homogenization. Firstly, we solve a neutron transport benchmark problem simulating fast critical assemblies composed of thin material plates with a discrete ordinates transport solver. The discrete ordinates transport solutions agree well with the Monte Carlo reference solutions; hence, we confirm the validity of the deterministic transport calculations for the sodium void reactivity worth of lattice-heterogeneous critical assemblies. Thereafter, the existing experimental data are calculated without whole-lattice homogenization. The result suggests that the lattice homogenization results in the overestimation of the leakage component of sodium void reactivity worth when the leakage component parallel to plate boundaries is dominant. Utilizing the numerical method without whole-lattice homogenization and the nuclear data JENDL-3.3, numerical solutions agree with the experimental data within 3 sigma of the experimental uncertainties.

Neutron transport benchmark problem proposal for fast critical assembly without homogenizations
Go Chiba, Kazuyuki Numata
ANNALS OF NUCLEAR ENERGY, 34, 6, 443, 448, PERGAMON-ELSEVIER SCIENCE LTD, Jun. 2007, [Peer-reviewed]
English, Scientific journal, In the present paper, we propose a neutron transport benchmark problem for fast critical assembly without homogenizations. With this problem, we can validate applicability of neutron transport codes when employed in highly heterogeneous fast critical assembly analyses. In addition, this benchmark problem can be used to validate homogenization procedures for slab lattices.
Detailed configurations of the cores and the lattices and cross-section data are provided in this paper. Reference solutions obtained with a Monte Carlo code are also provided. (C) 2007 Elsevier Ltd. All rights reserved.

Verification of homogenization in fast critical assembly analyses
Go Chiba
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 43, 11, 1395, 1405, ATOMIC ENERGY SOC JAPAN, Nov. 2006, [Peer-reviewed]
English, Scientific journal, In the present paper, homogenization procedures for fast critical assembly analyses are investigated.
Errors caused by homogenizations are evaluated by the exact perturbation theory. In order to obtain reference solutions, three-dimensional plate-wise transport calculations ire performed. It is found that the angular neutron flux along plate boundaries has a significant peak in the fission source energy range. To treat this angular dependence accurately, the double-Gaussian Chebyshev angular quadrature set with S-24 is applied.
It is shown that the difference between the heterogeneous leakage theory and the homogeneous theory is negligible, and that transport cross sections homogenized with neutron flux significantly underestimate neutron leakage. The error in criticality caused by a homogenization is estimated at about 0.1%, Delta k/kk' in a small fast critical assembly. In addition, the neutron leakage is overestimated by both leakage theories when sodium plates in fuel lattices are voided.

Improvement of moment-based probability table for resonance self-shielding calculation
Go Chiba, Hironobu Unesaki
ANNALS OF NUCLEAR ENERGY, 33, 13, 1141, 1146, PERGAMON-ELSEVIER SCIENCE LTD, Sep. 2006, [Peer-reviewed]
English, Scientific journal, In the present paper, an improved method has been proposed to produce a probability table needed for the resonance self-shielding calculations with the sub-group method. The proposed method is based on a relation between the effective cross section and the cross section moment, which is obtained from a numerical analysis. Using the proposed method, more accurate probability tables can be obtained with less number of the tabulated steps than the conventional method. This enables us to reduce computation time and computer memory storage for the sub-group calculations. (c) 2006 Elsevier Ltd. All rights reserved.

Overestimation in parallel component of neutron leakage observed in sodium void reactivity worth calculation for fast critical assemblies
Go Chiba
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 43, 8, 946, 949, ATOMIC ENERGY SOC JAPAN, Aug. 2006, [Peer-reviewed]
English, Scientific journal

Development of a fine and ultra-fine group cell calculation code SLAROM-UF for fast reactor analyses
Taira Hazama, Go Chiba, Kazuteru Sugino
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 43, 8, 908, 918, ATOMIC ENERGY SOC JAPAN, Aug. 2006, [Peer-reviewed]
English, Scientific journal, A cell calculation code SLAROM-UF has been developed for fast reactor analyses to produce effective cross sections with high accuracy in practical computing time, taking full advantage of fine and ultra-fine group calculation schemes.
The fine group calculation covers the whole energy range in a maximum of 900-group structure. The structure is finer above 52.5 keV with a minimum lethargy width of 0.008. The ultra-fine group calculation solves the slowing down equation below 52.5 keV to treat resonance structures directly and precisely including resonance interference effects. Effective cross sections obtained in the two calculations are combined to produce effective cross sections over the entire energy range.
Calculation accuracy and improvements from conventional 70-group cell calculation results were investigated through comparisons with reference values obtained with continuous energy Monte Carlo calculations. It was confirmed that SLAROM-UF reduces the difference in k-infinity from 0.15 to 0.01% for a JOYO MK-I fuel subassembly lattice cell calculation, and from -0.21% to less than a statistical uncertainty of the reference calculation of 0.03% for a ZPPR-10A core criticality calculation.

Revision and application of the covariance data processing code, ERRORJ
Go Chiba, Makoto Ishikawa
AIP Conference Proceedings, 769, 468, 471, 24 May 2005, [Peer-reviewed]
English, International conference proceedings, ERRORJ is the only code that can process the covariance data of the Reich-Moore resolved resonance parameters and the unresolved resonance parameters in the world. Now, the new version, version 2.2, has been developed and is released with improved reliability. In the present paper, details of the upgrade and a result of a validation test with the SAMMY code are described. Covariance data contained in ENDF/B, JEF(F), and JENDL are processed. Large differences are observed in the covariance between these nuclear data files. © 2005 American Institute of Physics.

Criticality analyses for fast neutron systems sensitive to iron
Go Chiba
Transactions of the Atomic Energy Society of Japan, 4, 1, 66, 76, Atomic Energy Society of Japan, 2005
Japanese, Scientific journal, Recently, it has been reported that the conventional calculation method for fast neutron systems loses its validity when it is applied to fast systems that are significantly sensitive to iron cross section. In the present paper, a newly developed cell calculation code, SLAROM-UF, has been applied to calculations for such systems. SLAROM-UF utilizes the ultra-fine energy group library below 50 keV and the 900-group library to estimate the self-shielding effect caused by resonances of heavy nuclides and wide resonances of structural materials, respectively. When SLAROM-UF with 900-group library was applied to cell calculation and core calculation was performed in a properly adopted 220-group structure, discrepancy of multiplication factor from the continuous energy Monte-Carlo calculation was reduced from 2.0 to 0.4%Δk. Large dependency on energy group used for core calculation is observed in JOYO MK-III. It is caused by "the fuel-reflector interface effect" which is recently discussed as a problem for calculation of fast neutron systems.

JOYO MK-III performance test at low power and its analysis
Gou Chiba, Kenji Yokoyama, Shigetaka Maeda, Takashi Sekine
Proceedings of the PHYSOR 2004: The Physics of Fuel Cycles and Advanced Nuclear Systems - Global Developments, 263, 273, Dec. 2004, [Peer-reviewed]
English, International conference proceedings, Performance test at low power has been carried out in the experimental fast reactor JOYO for the upgraded MK-III core. In the test, several neutronics characteristics, such as the control rod worth, the control rod shadowing effect, the excess reactivity and the isothermal temperature coefficient have been measured. For the analysis, a deterministic standard calculation method developed by Japan Nuclear Cycle Development Institute was used. Calculated values agreed well with the experimental ones within 0.55% Δk/kk' in the excess reactivity, 4% in the control rod worth and 3% in the isothermal temperature coefficient.

高速炉炉心解析における中性子非等方散乱と実効断面積の中性子束角度依存性の取扱いの臨界性への効果
千葉豪
日本原子力学会和文論文誌, 3, 2, 200, 207, 25 Jun. 2004
Japanese

Effect of neutron anisotropie scattering and treatment of angular dependency of neutron flux in effective cross section on criticality in fast reactor analysis
Gou Chiba
Transactions of the Atomic Energy Society of Japan, 3, 2, 200, 207, Atomic Energy Society of Japan, 2004
Japanese, Scientific journal, Numerical tests were performed about an effect of neutron anisotropic scattering and treatment of angular dependency of neutron flux in effective cross section on criticality based on previous researches. Three approximations described on a previous report were compared to each other in both one-dimensional slab model and two-dimensional cylindrical model. As a result, it was found that the transport approximation, which has been conventionally used in fast reactor analyses in Japan, has a good accuracy in criticality analyses of typical fast reactors. However, the transport approximation is not enough to calculate accurately cores which have fuel-reflector boundary. Therefore it is desirable that the extended transport approximation with higher order of anisotropic scattering is used in the analyses for such cores.

A combined method to evaluate the resonance self shielding effect in power fast reactor fuel assembly calculation
G Chiba
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 40, 7, 537, 543, ATOMIC ENERGY SOC JAPAN, Jul. 2003, [Peer-reviewed]
English, Scientific journal, A combined method for evaluating the resonance self shielding effect in a two-dimensional power fast reactor fuel assembly is proposed. This method uses the sub-group method to evaluate the self shielding effect of heterogeneous cell and collision probability method in the ultra-fine energy groups to deal with the resonance interference effect between different resonant nuclides.
In the present paper, a comparison between the table look-up method and the sub-group method is carried out and it is shown that the latter is superior to the former in a view of evaluating the resonance self shielding effect. These methods have a common defect that it is impossible to treat the resonance interference effect between different resonant nuclides. It can be overcome by using "a correction factor of the resonance interference effect" obtained by the collision probability calculation in the ultra-fine energy groups in a single fuel pin cell model. The microscopic effective cross section obtained by this proposed method agrees well with that by continuous energy Monte Carlo code within 5% relative difference above 100 eV. The k(infinity) value is affected about 0.1%Deltak/k by the use of the correction factor.

The revision of nuclear constant set for fast reactor, JFS-3-J3.2
Gou Chiba, Taira Hazama, Makoto Ishikawa
Transactions of the Atomic Energy Society of Japan, 1, 4, 335, 340, 2002
English, Scientific journal, The fast reactor group constant set JFS-3-J3.2 based on the evaluated nuclear data library JENDL-3.2 has been widely used in fast reactor analysis. However, it was recently found that there were errors in the process of making the group constant and they were revised. This set is called JFS-3-J3.2R. In this report, effects of the errors on nuclear characteristics were evaluated by a comparison with a new reactor group constant set, JFS-3-J3.2R. This report shows that the errors mainly affect removal cross section and distort neutron spectrum. As a result nuclear characteristics, such as sample Doppler reactivity and reaction rate in a blanket region, are significantly affected. However, it is also shown that other characteristics, such as criticality and sodium void reactivity, are not affected because the effects of errors are canceled out as a total integrated result. © 2002, Atomic Energy Society of Japan. All rights reserved.

Development of the hierarchical domain decomposition boundary element method for solving the three-dimensional multiregion neutron diffusion equations
G Chiba, M Tsuji, Y Shimazu
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 38, 8, 664, 673, TAYLOR & FRANCIS LTD, Aug. 2001, [Peer-reviewed]
English, Scientific journal, A hierarchical domain decomposition boundary element method (HDD-BEM) that was developed to solve a two-dimensional neutron diffusion equation has been modified to deal with three-dimensional problems. In the HDD-BEM, the domain is decomposed into homogeneous regions. The boundary conditions on the common inner boundaries between decomposed regions and the neutron multiplication factor are initially assumed. With these assumptions, the neutron diffusion equations defined in decomposed homogeneous regions can be solved respectively by applying the boundary element method. This part corresponds to the "lower level" calculations. At the "higher level" calculations, the assumed values, the inner boundary conditions and the neutron multiplication factor, are modified so as to satisfy the continuity conditions for the neutron flux and the neutron currents on the inner boundaries. These procedures of the lower and higher levels are executed alternately and iteratively until the continuity conditions are satisfied within a convergence tolerance. With the hierarchical domain decomposition, it is possible to deal with problems composing a large number of regions, something that has been difficult with the conventional BEM. In this paper, it is showed that a three-dimensional problem even with 722 regions can be solved with a fine accuracy and an acceptable computation time.

A hierarchical domain decomposition boundary element method with a higher order polynomial expansion for solving 2-D multiregion neutron diffusion equations
G Chiba, M Tsuji, Y Shimazu
ANNALS OF NUCLEAR ENERGY, 28, 9, 895, 912, PERGAMON-ELSEVIER SCIENCE LTD, Jun. 2001, [Peer-reviewed]
English, Scientific journal, A hierarchical domain decomposition boundary element method (HDD-BEM) for solving the multiregion neutron diffusion equation (NDE) has been developed to reduce computation time. The boundary integral equations derived from NDEs defined in homogeneous subregions are discretized with higher order boundary elements. The neutron flux and the neutron currents on boundary elements are expanded by quadratic or cubic polynomials. This expansion allows a large decrease in the number of unknown variables compared with the conventional HDD-BEM with constant boundary elements and reduces the computation time greatly. To obtain high accuracy with a small number of unknowns it is important to assign suitable nodal points on the non-conforming boundary elements. Guidelines for the assignment of nodal points is presented through numerical analysis. The HDD-BEM with higher order boundary elements calculates at least 5 times faster than the conventional HDD-BEM with constant boundary elements and 30 times faster than the finite difference method. The improvements in computation time will enable an extension of the scope of application to a wider variety of problems in reactor analysis. (C) 2001 Elsevier Science Ltd. All rights reserved.

High-speed parallel solution of the neutron diffusion equation with the hierarchical domain decomposition boundary element method incorporating parallel communications
M Tsuji, G Chiba
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 37, 5, 477, 485, TAYLOR & FRANCIS LTD, May 2000, [Peer-reviewed]
English, Scientific journal, A hierarchical domain decomposition boundary element method (HDD-BEM) for solving the multiregion neutron diffusion equation (NDE) has been fully parallelized, both for numerical computations and for data communications, to accomplish a high parallel efficiency on distributed memory message passing parallel computers. Data exchanges between node processors that are repeated during iteration processes of HDD-BEM are implemented, without any intervention of the host processor that was used to supervise parallel processing in the conventional parallelized HDD-BEM (P-HDD-BEM). Thus, the parallel processing can be executed with only cooperative operations of node processors. The communication overhead was even the dominant time consuming part in the conventional P-HDD-BEM, and the parallelization efficiency decreased steeply with the increase of the number of processors. With the parallel data communication, the efficiency is affected only by the number of boundary elements assigned to decomposed subregions, and the communication overhead can be drastically reduced. This feature can be particularly advantageous in the analysis of three-dimensional problems where a large number of processors are required.
The proposed P-HDD-BEM offers a promising solution to the deterioration problem of parallel efficiency and opens a new path to parallel computations of NDEs on distributed memory message passing parallel computers.

Development of Nuclear Data Processing CodeFRENDY Version 2
Kenichi Tada, Akio Yamamoto, Tomohiro Endo, Go Chiba, Michitaka Ono, Masayuki Tojo, Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022), 107, 116, 15 May 2022, [Peer-reviewed], [International Magazine]
Nuclear data processing is an important interface between an evaluated nuclear data library and neutronics calculation codes. JAEA has been developed the new nuclear data processing code FRENDY from 2013. FRENDY version 1 generates the ACE files used for the continuous-energy Monte Carlo codes including PHITS, Solomon, Serpent, and MCNP and it was released as the open-source software under the 2-clause BSD license in 2019.After we released FRENDY version 1, many functions, e.g., the multi-group neutron cross-section library generation, the statistical uncertainty quantification of the probability tables for unresolved resonance cross-section, the perturbation of the ACE file, and the modification of the ENDF-6 formatted nuclear data file, are developed. We released FRENDY version 2including these functions. The present paper explains the overview of FRENDY and features of new functions implemented in FRENDY version 2., American Nuclear Society, English, Summary international conference

FRENDY/MG: a multi-group cross section generation module using ace pointwise cross sections
Akio Yamamoto, Tomohiro Endo, Basma Foad, Go Chiba, Kenichi Tada, Proc. M&C 2021, 710, 720, 03 Oct. 2021, [Peer-reviewed], [International Magazine]
A generation capability of neutron multi-group cross sections is being implemented to the FRENDY nuclear data processing code, as FRENDY/MG. FRENDY/MG generates neutron multi-group cross sections for deterministic core analysis codes considering an arbitrary energy group structure. Distinguished features of FRENDY/MG are 1) use of ACE pointwise cross sections as the source of nuclear data (no evaluated nuclear data file is directly used), 2) treatment of a compound material consisting of multiple nuclides to explicitly consider the resonance interference effect. Various verifications are being carried out through the comparison with the multi-group cross sections generated by NJOY., American Nuclear Society, English, Summary international conference

Multi-Group Cross Section Library Generation by FRENDY for Fast Reactor Neutronics Calculations
Go Chiba, Akio Yamamoto, Kenichi Tada, Tomohiro Endo, Transactions of the American Nuclear Society, 124, 556, 558, Jun. 2021, [Peer-reviewed], [International Magazine]
American Nuclear Society, English, Summary international conference

Verification of the Multi-Group Generation Capability of FRENDY Nuclear Data Processing Code for Recent Nuclear Data through Comparison of One-group Reaction Rates
Akio Yamamoto, Kenichi Tada, Go Chiba, Tomohiro Endo, Transactions of the American Nuclear Society, 124, 544, 547, Jun. 2021, [International Magazine]
American Nuclear Society, English, Summary international conference

Treatment of scattering sources in numerical calculations of the radiative transfer equation with discrete ordinates methods
清水亮輔, 千葉豪, 日本原子力学会春の年会予稿集(CD-ROM), 2021, 2021

Improvement of high-efficiency random sampling method by using control variates method and sensitivity (2) Application to burnup characteristics of BWR fuel assembly
木田拓実, 千葉豪, 日本原子力学会春の年会予稿集(CD-ROM), 2021, 2021

Study on Optimization of Fuel Composition for Reducing Sodium Void Reactivity in Sodium-Cooled Fast Reactor
大池宏弥, 遠藤知弘, 山本章夫, 千葉豪, 日本原子力学会春の年会予稿集(CD-ROM), 2021, 2021

Development of pseudo FP generation algorithm for nuclear fuel burnup calculations (1) Outline of the algorithm
柳原健人, 千葉豪, 日本原子力学会春の年会予稿集(CD-ROM), 2021, 2021

Development of pseudo FP generation algorithm for nuclear fuel burnup calculations (2) Numerical results
柳原健人, 千葉豪, 日本原子力学会春の年会予稿集(CD-ROM), 2021, 2021

原子力エネルギーシステム応用の観点から
千葉豪, 核データニュース(Web), 123, 34‐41 (WEB ONLY), Jun. 2019
Japanese

核データ積分テスト計算への実効部分空間法の適用 2)
今里大智, 千葉豪, 日本原子力学会春の年会予稿集(CD-ROM), 2019, ROMBUNNO.3J06, 04 Mar. 2019
Japanese

原子力エネルギーシステム応用の観点から
千葉豪, 日本原子力学会春の年会予稿集(CD-ROM), 2019, ROMBUNNO.3E_PL04, 04 Mar. 2019
Japanese

21世紀後半に向けた廃棄物管理の選択肢:Pu利用推進と環境負荷低減型地層処分に関する研究(8)MAリサイクルを想定した高速炉の廃棄物特性評価
小林優, 千葉豪, 川久保政洋, 朝野英一, 日本原子力学会春の年会予稿集(CD-ROM), 2019, ROMBUNNO.1B08, 04 Mar. 2019
Japanese

21世紀後半に向けた廃棄物管理の選択肢:Pu利用推進と環境負荷低減型地層処分に関する研究(7)分離核種の燃焼を念頭に置いたCBZコードシステムを用いた高速炉燃焼計算
千葉豪, 川久保政洋, 朝野英一, 日本原子力学会春の年会予稿集(CD-ROM), 2019, ROMBUNNO.1B07, 04 Mar. 2019
Japanese

感度係数による不確かさ評価結果を類似対称に用いた,より効果的なMCCV法の提案
二平舜介, 千葉豪, 日本原子力学会秋の大会予稿集(CD-ROM), 2018, ROMBUNNO.1M03, 20 Aug. 2018
Japanese

CBZコードシステムを用いた,福島第一原子力発電所内部の放射能インベントリ解析の詳細化
石井亮憲, 千葉豪, 日本原子力学会秋の大会予稿集(CD-ROM), 2018, ROMBUNNO.3I10, 20 Aug. 2018
Japanese

加速器駆動システムによる核変換処理の実現に向けた基礎研究(6)U‐Pbゾーン炉心を用いたADS実験
PYEON Cheol Ho, 山中正朗, 渡辺賢一, 相澤直人, 遠藤知弘, 千葉豪, VAN ROOIJEN Wilfred G, 日本原子力学会秋の大会予稿集(CD-ROM), 2018, ROMBUNNO.2M02, 20 Aug. 2018
Japanese

加速器駆動システムによる核変換処理の実現に向けた基礎研究(7)動特性パラメータの実験解析
山中正朗, 遠藤知弘, 千葉豪, VAN ROOIJEN Wilfred G, PYEON Cheol Ho, 日本原子力学会秋の大会予稿集(CD-ROM), 2018, ROMBUNNO.2M03, 20 Aug. 2018
Japanese

炉物理コードシステムCBZへのガラーキン求積法の実装
千葉豪, 藤井宏之, 日本原子力学会秋の大会予稿集(CD-ROM), 2018, ROMBUNNO.2M20, 20 Aug. 2018
Japanese

On prediction accuracy of neutronics parameters of accelerator-driven system
Go Chiba, Tomohiro Endo, Wilfred Van Rooijen, Cheol Ho Pyeon, EPJ Web of Conferences, 146, 13 Sep. 2017, [Peer-reviewed]
Nuclear data-induced uncertainty of neutronics parameters of one specific ADS design is quantified. The nuclear data adjustment method with available integral data is employed to reduce the uncertainties, and usefulness of these integral data is investigated. Numerical results reveal that the uncertainty reduction by the present nuclear data adjustment is insignificant and restrictive. Future perspecitives are also provided., EDP Sciences, English

多核種高除染性空気浄化システム開発による作業被曝低減化研究(8)フィルターエレメントに捕捉される放射能評価のための簡易ツールの開発
千葉豪, 石井亮憲, 小崎完, 奈良林直, 佐藤修彰, 日本原子力学会秋の大会予稿集(CD-ROM), 2017, ROMBUNNO.2F21, 29 Aug. 2017
Japanese

多核種高除染性空気浄化システム開発による作業被曝低減化研究(9)フィルターエレメントに捕捉される短半減期FPの放射能の評価
石井亮憲, 千葉豪, 小崎完, 奈良林直, 山本泰功, 日本原子力学会秋の大会予稿集(CD-ROM), 2017, ROMBUNNO.2F22, 29 Aug. 2017
Japanese

Implementation of Fuel Depletion Sensitivity Calculation Capability into a Deterministic Reactor Physics Code System CBZ for Accelerator-Driven Sub-System Multi-Cycle Burnup Calculations　　　　　　　　　　　　　　　
G. Chiba, W. F, G. van Rooijen, T. Endo, C. H. Pyeon, Proc. Int. Conf. Mathematics & Computational Methods Applied to Nucl. Sci. & Eng. (M&C2017), Jeju, Korea, Apr. 16-20, (2017)., Apr. 2017, [Peer-reviewed]
English, Summary international conference

Nuclear Data-Induced Uncertainty Quantification of Neutron Multiplication Factor and Prompt Neutron Decay Constant for Pb-Bi loaded ADS Benchmark Problems at KUCA　　　　　　　　　　　　　　　
T. Endo, G. Chiba, W. F, G. van Rooijen, M. Yamanaka, C. H. Pyeon, Proc. Int. Conf. Mathematics & Computational Methods Applied to Nucl. Sci. & Eng. (M&C2017), Jeju, Korea, Apr. 16-20, (2017)., Apr. 2017, [Peer-reviewed]
English, Summary international conference

Variance Reduction Factor Calculations for Neutronics Parameters of Accelerator-Driven System　　　　　　　　　　　　　　　
G. Chiba, W. F, G. van Rooijen, T. Endo, C. H. Pyeon, Proc. Int. Conf. on the Physic of Reactors (PHYSOR2016), Sun Valley, Idaho, May 1-5, (2016)., May 2016, [Peer-reviewed]
English, Summary international conference

Analysis and Interpretation of the KUCA ADS Benchmarks with Deterministic Analysis Codes　　　　　　　　　　　　　　　
W. F, G. van Rooijen, C. H. Pyeon, G. Chiba, T. Endo, Proc. Int. Conf. on the Physic of Reactors (PHYSOR2016), Sun Valley, Idaho, May 1-5, (2016)., May 2016, [Peer-reviewed]
English, Summary international conference

多核種高除染性空気浄化システム開発による作業被曝低減化研究(1)全体計画
奈良林直, 千葉豪, 小崎完, 増田隆夫, 中坂佑太, 佐藤修彰, 秋山大輔, 日本原子力学会春の年会予稿集(CD-ROM), 2016, ROMBUNNO.1D20, 16 Mar. 2016
Japanese

Nurturing Problem-Finding Skills in Graduate Students through Problem Based Learning Approaches
Ankit A. Ravankar, Shotaro Imai, Michiyo Shimamura, Go Chiba, Taichi Takasuka, Yasuhiro Yamanaka, PROCEEDINGS 2016 5TH IIAI INTERNATIONAL CONGRESS ON ADVANCED APPLIED INFORMATICS IIAI-AAI 2016, 542, 546, 2016, [Peer-reviewed]
The present university education system has been designed to make graduate students good problem solvers such that they can contribute to society through the skills acquired in the graduate schools. With focus on developing critical thinking and reasoning to solve local and global problems such programs have gained immense popularity among teachers and professors in schools and universities and termed as "Problem-Based-Learning" (PBL). However, there has been very few programs that encourage students to become better Problem finders. This study is based on our own experience of teaching problem-finding as an important skill for a special graduate program in university education and our findings on its implication on students ability to comprehend real world problems., IEEE, English

Discussion on a method of Team-Based-Learning style lecture for graduate students in a research university
Shotaro Imai, Ankit A. Ravankar, Michiyo Shimamura, Taichi E. Takasuka, Go Chiba, Yasuhiro Yamanaka, PROCEEDINGS 2016 5TH IIAI INTERNATIONAL CONGRESS ON ADVANCED APPLIED INFORMATICS IIAI-AAI 2016, 537, 541, 2016, [Peer-reviewed]
The new graduate school education program Nitobe School was launched in 2015 as one of main education projects of "Top Global University Project" in Hokkaido University. This new graduate program is aimed to supply currently undermined skill-set to students from different home graduate or professional school. The scope of the Nitobe School program is to foster students to be a global leader who might contribute not only for local societies but also internationally by Team-Based-Learning (TBL) and Project-Based-Learning (PBL) methods. Since students have different expertise including the area of STEM and humanity, teachers cannot guide classes with their field, rather teachers need to educate students using a general topics. This general education program apparently requires different approaches from the conventional one. After completion of the first year of Nitobe School, we took place the survey that asked students how they felt about our new teaching methods. Overall, students appreciated with the new methods that we carried out, but there are some spaces for further improvement. In this report, we suggest potential ways to improve the Nitobe School program, which will be able to enhance further educating students, and at the same time educate teachers toward the TBL and PBL., IEEE, English

B212 Improvement of Two-Phase Flow Stability for Filtered Containment Venting System Using Silver Zeolite
SHOBUKAWA Hideyuki, NARABAYASHI Tadashi, CHIBA Go, Sato Masanobu, National Symposium on Power and Energy Systems, 2015, 20, 221, 222, 07 Jun. 2015
We have conducted a series of experiments in order to investigate and improve the performance of the advanced filtered containment venting system: a visualization experiment to examine two-phase flow stability in a vertical tube, a superheated steam experiment with multi-layer orifices and a clogging experiment with metal filter and simulated grains. In the first experiment, several specific phenomena, such as chugging, CCFL and geysering, have been observed. Generation of superheated steam has been confirmed through the second experiment. Regarding the clogging experiment, whereas metal filter has not been clogged, accumulation of the simulated grains has been observed., The Japan Society of Mechanical Engineers, Japanese

B135 Evaluation on Decay Heat Removal Capability of Isolation Condenser for Depressurization iand Cooling System
NARABAYASHI Tadashi, SHIMOE Tomohiro, CHIBA Go, TSUJI Masashi, SAKASHITA Hiroto, National Symposium on Power and Energy Systems, 2014, 19, 73, 74, 25 Jun. 2014
In this study, we aimed to quantitatively evaluate cooling capacity of isolation condenser (IC) for core cooling in Fukushima Daiichi Nuclear Plant Unit 1. In order to evaluate cooling capacity, decay heat emitted from fuels in reactor core was calculated. We used the data of nuclear fuel and rated thermal power which is released by TEPCO. We analyzed IC with TRAC code. We made an experiment in order to check IC's input data. Because analysis data is close to experimental data, we made sure validity of input data. We analyzed IC under conditions which are recorded when IC started up and tsunami rushed toward plant. As a result of analysis, we confirmed that cooling capacity exceed decay heat under both conditions. In conclusion, IC had a capability to prevent core meltdown accident in Fukushima Daiichi Nuclear Plant Unit 1., The Japan Society of Mechanical Engineers, Japanese

B225 Molten Core Concrete Interaction and Development of Core Catcher
NARABAYASHI Tadashi, MIYAWAKI Daichi, SYLWESTER Marta Ziemnicka, CHIBA Go, TSUJI Masashi, National Symposium on Power and Energy Systems, 2014, 19, 257, 258, 25 Jun. 2014
Fukushima Daiichi NPP accident would be terminated, if sufficient accident countermeasures, such as water proof door, mobile power, etc. In case of Europe, it had already recommended to install core catcher from the lessons of Chernobyl Accidents. In this paper we introduce simplified core catch by using core catcher that used Basalt. Hokkaido University has tested the simplified core catcher, by using thermite process. We found that molten thermite debris and concrete interaction make concrete to erode on its surface. We also tested several ceramics such as SiC etc., to find suitable materials. But the ceramics was broken by thermal shock. Among these materials we recommend the Basalt for the material of the core catcher., The Japan Society of Mechanical Engineers, Japanese

トリウム装荷臨界実験に対する感度解析と不確かさ評価
田熊伸也, 千葉豪, 辻雅司, 奈良林直, 日本原子力学会春の年会予稿集(CD-ROM), 2014, ROMBUNNO.O50, 10 Mar. 2014
Japanese

ランダムサンプリング法を用いた断面積調整法および感度係数評価(3)感度係数評価
千葉豪, 辻雅司, 奈良林直, 渡辺友章, 遠藤知弘, 山本章夫, 日本原子力学会春の年会予稿集(CD-ROM), 2014, ROMBUNNO.O47, 10 Mar. 2014
Japanese

核種生成量に対する核データの感度解析
川本洋右, 千葉豪, 辻雅司, 奈良林直, 日本原子力学会春の年会予稿集(CD-ROM), 2014, ROMBUNNO.O49, 10 Mar. 2014
Japanese

Influences of Differences in Nuclear Data Libraries on Important Back-end Parameters
YOSHIDA SHOTA, CHIBA GO, TSUJI MASASHI, NARABAYASHI TADASHI, 日本原子力研究開発機構JAEA-Conf(Web), 2013-002, WEB ONLY 28, Oct. 2013
English

Validation of CBZ Code System for Post-Irradiation Examination Analysis and Sensitivity Analysis of (n,.GAMMA.) Branching Ratio
KAWAMOTO YOSUKE, CHIBA GO, TSUJI MASASHI, NARABAYASHI TADASHI, 日本原子力研究開発機構JAEA-Conf(Web), 2013-002, WEB ONLY 21, Oct. 2013
English

Validation of JENDL-4.0 and Future: Reactor Integral Test Working Group
CHIBA GO, 日本原子力研究開発機構JAEA-Conf(Web), 2013-002, WEB ONLY 12, Oct. 2013
English

高速炉蒸気発生器における伝熱管破損事象に関する研究(27)管群体系内の蒸気噴流可視化試験 第二報
加藤慶輔, 奈良林直, 辻雅司, 千葉豪, 大島宏之, 栗原成計, 内堀昭寛, 日本原子力学会秋の大会予稿集(CD-ROM), 2013, ROMBUNNO.J12, 20 Aug. 2013
Japanese

J‐PARC/MLF/ANNRI NaI(Tl)スペクトロメータ及びGeスペクトロメータを用いたGd‐157の中性子捕獲断面積の測定
木村敦, 廣瀬健太郎, 中村詔司, 原かおる, 原田秀郎, 千葉豪, 片渕竜也, 寺田和司, 水本元治, 井頭政之, 日本原子力学会秋の大会予稿集(CD-ROM), 2013, ROMBUNNO.C02, 20 Aug. 2013
Japanese

核特性パラメータの不確かさ評価における共鳴自己遮蔽効果
千葉豪, 辻雅司, 奈良林直, 日本原子力学会秋の大会予稿集(CD-ROM), 2013, ROMBUNNO.J21, 20 Aug. 2013
Japanese

特異値分解法による簡易燃焼チェーン自動作成ツールの開発
梶原孝則, 辻雅司, 千葉豪, 奈良林直, 大岡靖典, 牛尾直史, 日本原子力学会秋の大会予稿集(CD-ROM), 2013, ROMBUNNO.J28, 20 Aug. 2013
Japanese

原子力発電所の各種漏洩事象評価手法の高度化に関する研究(1)漏洩評価法の高度化研究の概要
岡田健志, 奈良林直, 辻雅司, 千葉豪, 尾形悠斗, 宮野廣, 松本昌昭, 日本原子力学会秋の大会予稿集(CD-ROM), 2013, ROMBUNNO.K49, 20 Aug. 2013
Japanese

ゼオライトを用いた高除染性フィルタードベントシステムの開発(5)フィルタードベントシステムの作動範囲とパラメータ試験
藤井康弘, 奈良林直, 佐藤修彰, 辻雅司, 千葉豪, 川本洋右, 日本原子力学会秋の大会予稿集(CD-ROM), 2013, ROMBUNNO.H50, 20 Aug. 2013
Japanese

非常用復水器を用いた過酷事故緩和に関する研究 高圧可視化実験装置を用いた実験
下江知弘, 奈良林直, 辻雅司, 千葉豪, 日本原子力学会秋の大会予稿集(CD-ROM), 2013, ROMBUNNO.H41, 20 Aug. 2013
Japanese

ゼオライトを用いた高除染性フィルタードベントシステムの開発(4)格納容器のFP保持機能を活かしたフィルタードベントシステム
奈良林直, 藤井康弘, 佐藤修彰, 辻雅司, 千葉豪, 日本原子力学会秋の大会予稿集(CD-ROM), 2013, ROMBUNNO.H49, 20 Aug. 2013
Japanese

原子力発電所の各種漏洩事象評価手法の高度化に関する研究(2)配管からの微小漏洩評価手法の開発
尾形悠斗, 奈良林直, 辻雅司, 千葉豪, 岡田健志, 松本昌昭, 宮野廣, 日本原子力学会秋の大会予稿集(CD-ROM), 2013, ROMBUNNO.K50, 20 Aug. 2013
Japanese

E113 Visualization Test of Filtered Vent and Evaluation of Source Term
FUJII Yasuhiro, KAWAMOTO Yosuke, NARABAYASHI Tadashi, TSUJI Masashi, CHIBA Go, National Symposium on Power and Energy Systems, 2013, 18, 151, 152, 19 Jun. 2013
This paper describes how to write a manuscript of the 18th PES Proceedings. In the Fukushima Daiichi Nuclear Power Plant accident, a great deal of radioactive material was released to surrounding environment After the accident, the Filtered Containment Venting System (FCVS) that can remove radioactive material from steam in containment vessel has attracted a great deal of attentions. However, there is few knowledge of filtered vent in Japan because almost filtered vent were developed in Europe after Chernobyl nuclear power plant accident. Therefore, our purpose is getting the knowledge of filtered vent through experiments that simulate various filter vents in Europe. Also using the knowledge gotten through experiment, we aim at developing high efficiency filtered containment venting system using zeolite., The Japan Society of Mechanical Engineers, Japanese

ND2013核データ国際会議報告(3)炉物理関連(積分実験)
福島昌弘, 千葉豪, 核データニュース(Web), 105, WEB ONLY 37-39, Jun. 2013
Japanese

高速炉蒸気発生器における伝熱管破損事象に関する研究(23)管群体系内の蒸気噴流可視化試験
加藤慶輔, 新栄邦彦, 奈良林直, 辻雅司, 千葉豪, 大島宏之, 栗原成計, 内堀昭寛, 日本原子力学会春の年会予稿集(CD-ROM), 2013, ROMBUNNO.N16, 11 Mar. 2013
Japanese

ゼオライトを用いた高除染性フィルタードベントシステムの開発(1)フィルタードベントシステムの基本構成の検討と放射性核種吸着実験
奈良林直, 藤井康弘, 佐藤修彰, 辻雅司, 千葉豪, 日本原子力学会春の年会予稿集(CD-ROM), 2013, ROMBUNNO.M08, 11 Mar. 2013
Japanese

一般化摂動論を用いた燃焼感度の計算
千葉豪, 辻雅司, 奈良林直, 日本原子力学会春の年会予稿集(CD-ROM), 2013, ROMBUNNO.H46, 11 Mar. 2013
Japanese

ゼオライトを用いた高除染性フィルタードベントシステムの開発(2)フィルター内流動可視化試験
藤井康弘, 奈良林直, 佐藤修彰, 辻雅司, 千葉豪, 日本原子力学会春の年会予稿集(CD-ROM), 2013, ROMBUNNO.M09, 11 Mar. 2013
Japanese

ゼオライトを用いた高除染性フィルタードベントシステムの開発(3)多核種解析法によるソースタームと崩壊熱計算
川本洋右, 藤井康弘, 奈良林直, 佐藤修彰, 辻雅司, 千葉豪, 日本原子力学会春の年会予稿集(CD-ROM), 2013, ROMBUNNO.M10, 11 Mar. 2013
Japanese

動特性同定法によるもんじゅの反応度温度係数の測定
佐久間渉, 辻雅司, 奈良林直, 千葉豪, KURRI KR (CD) (CD-ROM), 39, ROMBUNNO.8, 2013
Japanese

過酷事故緩和に対する深層防護の強化の提案(3)可視化試験結果
藤井康弘, 奈良林直, 石田和輝, 千葉豪, 日本原子力学会秋の大会予稿集(CD-ROM), 2012, ROMBUNNO.M39, 346, 03 Sep. 2012
蒸気を用いたフィルター付ベントシステムの可視化試験試験を実施して、フィルター部の流動特性や圧力損失の測定、ミストセパレータの液滴除去特性について測定した。また、ゼオライトのセシウム吸着特性についても測定し、ゼオライトを用いたフィルタードベントシステムの放射性物質の除去特性についても基礎実験を行った。, Atomic Energy Society of Japan, Japanese

一点炉随伴動特性方程式を用いた原子炉ペリオドの感度解析
千葉豪, 辻雅司, 奈良林直, 日本原子力学会秋の大会予稿集(CD-ROM), 2012, ROMBUNNO.Q49, 173, 03 Sep. 2012
遅発中性子先行核を陽に扱った一点炉随伴動特性方程式を用いて、原子炉ペリオドの核データに対する感度解析を行った。, Atomic Energy Society of Japan, Japanese

詳細燃焼チェーンを用いた高速炉全炉心計算
吉田将太, 千葉豪, 奈良林直, 辻雅司, 日本原子力学会秋の大会予稿集(CD-ROM), 2012, ROMBUNNO.Q17, 141, 03 Sep. 2012
バックエンド諸量の正確な評価のために核分裂生成物核種200程度を陽に扱った高速炉全炉心計算を行った。また、簡略化したモデルによる計算結果の誤差を定量的に評価した。, Atomic Energy Society of Japan, Japanese

Analysis of Radioactivity of the Ground Surface Measured at the Fukushima-Daiichi Nuclear Power Plant Site
CHIBA GO, 日本原子力研究開発機構JAEA-Conf(Web), 2012-001, WEB ONLY 23, Jul. 2012
English

Development of a Standard Data Base for FBR Core Design (XIV)-Analyses of Extensive FBR Core Characteristics Based on JENDL-4.0-
SUGINO KAZUTERU, ISHIKAWA MAKOTO, NUMATA KAZUYUKI, NUMATA KAZUYUKI, IWAI TAKEHIKO, IWAI TAKEHIKO, JIN TOMOYUKI, JIN TOMOYUKI, NAGAYA YASUNOBU, HAZAMA TAIRA, CHIBA GO, CHIBA GO, YOKOYAMA KENJI, KUGO TERUHIKO, 日本原子力研究開発機構JAEA-Research(Web), 2012-013, WEB ONLY, Jul. 2012
Japanese

海外の過酷事故対策の取り組みとフィルター付ベント
奈良林直, 杉山憲一郎, 石田和輝, 辻雅司, 千葉豪, 岡本孝司, 水町渉, 日本機械学会動力・エネルギー技術シンポジウム講演論文集, 17th, 17, 335, 338, 20 Jun. 2012
Fukushima Daiichi NPP accident would be terminated, if sufficient examination leads to install countermeasures, such as water proof door, mobile power, etc. In case of US, Diablo Canyon NPP in Florida, they have a water proof hatch type door and snorkel piping for the seawater pump motor. In case of Europe, it had already installed the heat removal system and filtered containment venting system from the lessons of TMI and Chernobyl Accidents. Decay heat removal system and CV spray cooling system with FCVS are ensured by using mobile generators and heat exchangers to keep the ultimate heat sink even in any natural disaster, such as large earthquake, big tsunami, sudden flooding etc. In this paper we introduce such technology and fundamental test for the filtered venting system, conducted by Hokkaido University., The Japan Society of Mechanical Engineers, Japanese

福島第1原子力発電所の非常用復水器の除熱能力評価
下江知弘, 奈良林直, 辻雅司, 千葉豪, 日本機械学会動力・エネルギー技術シンポジウム講演論文集, 17th, 17, 333, 334, 20 Jun. 2012
In this study, we aimed to quantitatively evaluate cooling capacity of isolation condenser (IC) for core cooling in Fukushima Daiichi Nuclear Plant Unit 1. In order to evaluate cooling capacity, we calculated decay heat which was emitted from core. We used the data of nuclear fuel and rated thermal power which is released by TEPCO. Analysis was conducted with TRAC code. We made an experiment in order to check input data for IC. Because analysis data was close to experimental data, we made sure of validity of input data. We analyzed IC under conditions which were recorded when IC started up and tsunami rushed toward plant. As a result of analysis, we confirmed that cooling capacity exceed decay heat under both conditions. In conclusion, IC had a capacity to prevent core meltdown accident in Fukushima Daiichi Nuclear Plant Unit 1., The Japan Society of Mechanical Engineers, Japanese

惑星間航行用原子力推進宇宙船の放射線遮蔽と居住性評価
岡田健志, 奈良林直, 辻雅司, 千葉豪, 日本機械学会動力・エネルギー技術シンポジウム講演論文集, 17th, 17, 327, 328, 20 Jun. 2012
The space development is our big dream. One of this is foray into deep space. And so we must have too large amount of power because missions in space is highly-developed and protracted, and in addition, devices are complicated and big with the space development. Nuclear power generation is good choice to these problems since it has such good points as big electric-generating capacity in the proximity of scale and weight of the facility. There our laboratory has researched about a nuclear spaceship. Especially this paper deals with living space in a nuclear spaceship. For example space has big difference in temperature and nuclear reactor releases radiation. The crew is needed to protect from such danger. First we did reproductive experiment of big difference in temperature in space as a basic experiment of heat insulation and researched the possibility of insulation by insulating material of urethane foam. Next we tried a design of shielding system by using a calculation code., The Japan Society of Mechanical Engineers, Japanese

過酷事故緩和に対する深層防護の強化の提案(2)フィルター付ベントシステムの基礎試験
奈良林直, 杉山憲一郎, 石田和輝, 水谷一貴, 千葉豪, 日本原子力学会春の年会予稿集(CD-ROM), 2012, ROMBUNNO.G40, 02 Mar. 2012
Japanese

多様な原子力システム間のバックエンド諸量の定量的比較
吉田将太, 千葉豪, 奈良林直, 辻雅司, 日本原子力学会北海道支部研究発表会講演要旨集, 30th, 19, 20, 2012
Japanese

動特性同定法を用いた高速増殖炉もんじゅの反応度温度係数の測定
佐久間渉, 辻雅司, 奈良林直, 千葉豪, 日本原子力学会北海道支部研究発表会講演要旨集, 30th, 21, 22, 2012
Japanese

Uncertainty Quantification of Doppler Coefficient for MONJU
CHIBA GO, CHIBA GO, HAZAMA TAIRA, KINJO HIDEHITO, KINJO HIDEHITO, NISHI HIROSHI, SUZUKI TAKAYUKI, 日本原子力研究開発機構JAEA-Research(Web), 2011, 2011-034, WEB ONLY, 2,1-42, Dec. 2011
日本原子力研究開発機構, Japanese

Uncertainty Quantification of Doppler Coefficient for MONJU
千葉 豪, 羽様 平, 金城 秀人, JAEA-research, 2011, 34, 巻頭1, 2,1-42, Dec. 2011
日本原子力研究開発機構, Japanese

Sensitivity Analysis for Higher Order Legendre Coefficients of Elastic Scattering Matrices
CHIBA Go, 日本原子力研究開発機構JAEA-Conf(Web), 2011-002, WEB ONLY 38, Sep. 2011
English

Sensitivity Analysis for Curium Isotope Concentrations of Light Water Reactor Mixed-Oxide Burned Fuel
CHIBA Go, ISHIKAWA Makoto, 日本原子力研究開発機構JAEA-Conf(Web), 2011-002, WEB ONLY 37, Sep. 2011
English

Improved Derivation of Multigroup Effective Cross Section for Heterogeneous System by Equivalence Theory
A. Yamamoto, T. Endo, G. Chiba, Nucl. Sci. Eng, 168, 75-92, 92, Jun. 2011, [Peer-reviewed]
English, Report scientific journal

On the cell homogenization for a lattice composed of asymmetric unit cells
CHIBA Go,VAN, ROOIJEN W.F.G, 日本原子力学会春の年会予稿集(CD-ROM), 2011, ROMBUNNO.F21, 11 Mar. 2011
English

Diffusion coefficients for asymmetric unit cells
VAN ROOIJEN, W.F.G, CHIBA Go, 日本原子力学会春の年会予稿集(CD-ROM), 2011, ROMBUNNO.F20, 11 Mar. 2011
English

Study on Calculation Methods for the Effective Delayed Neutron Fraction
IRWANTO Dwi, IRWANTO Dwi, CHIBA Go, NAGAYA Yasunobu, OBARA Toru, OBARA Toru, 日本原子力研究開発機構JAEA-Research(Web), 2010, 2010-061, WEB ONLY, 28,巻頭1-2, Mar. 2011
日本原子力研究開発機構, English

Development of the Next Generation Reactor Analysis Code System, MARBLE
YOKOYAMA KENJI, TATSUMI MASAHIRO, TATSUMI MASAHIRO, HIRAI YASUSHI, HIRAI YASUSHI, HYODO HIDEAKI, HYODO HIDEAKI, NUMATA KAZUYUKI, NUMATA KAZUYUKI, IWAI TAKEHIKO, IWAI TAKEHIKO, JIN TOMOYUKI, JIN TOMOYUKI, HAZAMA TAIRA, NAGAYA YASUNOBU, CHIBA GO, KUGO TERUHIKO, ISHIKAWA MAKOTO, 日本原子力研究開発機構JAEA-Data/Code(Web), 2010-030, WEB ONLY, Mar. 2011
Japanese

Study on Calculation Methods for the Effective Delayed Neutron Fraction
IRWANTO Dwi, 千葉 豪, 長家 康展, JAEA-research, 2010, 61, 1, 28,巻頭1-2, Mar. 2011
日本原子力研究開発機構, English

金属ウラン・ジルコニウムマトリックス燃料を装荷した小型PWRに関する研究
高田祐太, 奈良林直, 辻雅司, 千葉豪, 島津洋一郎, 日本原子力学会北海道支部研究発表会講演要旨集, 29th, 17, 18, 2011
Japanese

福島第一原子力発電所構内土壌の放射能データの解析
千葉豪, 杉山憲一郎, 片倉純一, 日本原子力学会北海道支部研究発表会講演要旨集, 29th, 31, 32, 2011
Japanese

Evaluation of Gamma-Ray Transport Effect on Heating Distribution in a Fast Reactor Fuel Subassembly
CHIBA GO, 日本原子力研究開発機構JAEA-Research(Web), 2010-045, WEB ONLY, Jan. 2011
Japanese

Analysis of MOZART Critical Experiment Using the IRPhEP Handbook Data
千葉 豪, JAEA-research, 2010, 42, 1, 92,巻頭1-2, Jan. 2011
日本原子力研究開発機構, Japanese

Evaluation of Gamma-Ray Transport Effect on Heating Distribution in a Fast Reactor Fuel Subassembly
千葉 豪, JAEA-research, 2010, 45, 1, 24,巻頭1-2, Jan. 2011
日本原子力研究開発機構, Japanese

Uncertainty Quantification of Lumped Fission Product Cross Section for Fast Reactor Application
CHIBA Go, 日本原子力研究開発機構JAEA-Conf(Web), 2010-005, WEB ONLY 9, Dec. 2010
English

Plan of the JENDL-4 Benchmark for Fast Reactors
SUGINO Kazuteru, CHIBA Go, 日本原子力研究開発機構JAEA-Conf(Web), 2010-005, WEB ONLY 7, Dec. 2010
English

Analysis of MOZART Critical Experiment Using the IRPhEP Handbook Data
CHIBA GO, 日本原子力研究開発機構JAEA-Research(Web), 2010-042, WEB ONLY, Dec. 2010
Japanese

原子力開発のための中性子核反応データベース 評価済み核データライブラリーJENDL‐4.0の完成
柴田恵一, 岩本修, 千葉豪, ΑΤΟΜΟΣ, 52, 12, 801, 805, 01 Dec. 2010
日本原子力学会, Japanese

JENDL-4.0 : A Database on Neutron-induced Reactions for Nuclear Science and Engineering
SHIBATA Keiichi, IWAMOTO Osamu, CHIBA Go, Ατομοσ, 52, 12, 801, 805, 01 Dec. 2010
日本原子力学会, Japanese

Development of Gamma Cross Section Library for Fast Reactor Gamma Heating Calculations
CHIBA GO, HAZAMA TAIRA, SUZUKI TAKAYUKI, 日本原子力研究開発機構JAEA-Data/Code(Web), 2010-013, WEB ONLY, Sep. 2010
Japanese

Calculation of diffusion coefficients for voided lattices with the method of characteristics: II: numerical results
CHIBA Go,VAN, ROOIJEN W.F.G, 日本原子力学会秋の大会予稿集(CD-ROM), 2010, ROMBUNNO.Q36, 27 Aug. 2010
English

Calculation of diffusion coefficients for voided lattices with the method of characteristics: I: theory
VAN ROOIJEN, W.F.G, CHIBA Go, 日本原子力学会秋の大会予稿集(CD-ROM), 2010, ROMBUNNO.Q35, 27 Aug. 2010
English

二項有理式近似を用いたToneの手法の改良
山本章夫, 遠藤知弘, 千葉豪, 日本原子力学会秋の大会予稿集(CD-ROM), 2010, ROMBUNNO.Q42, 27 Aug. 2010
Japanese

べき乗法の収束残差からの高次固有値の推定
中瀬正彦, 千葉豪, 齊藤正樹, 日本原子力学会春の年会予稿集(CD-ROM), 2010, ROMBUNNO.G19, 09 Mar. 2010
Japanese

高速炉用オブジェクト統合型解析システムの研究開発―(11)次世代炉物理解析システムMARBLE1.0の完成―
横山賢治, 巽雅洋, 平井康志, 兵頭秀昭, 沼田一幸, 岩井武彦, 神智之, 羽様平, 長家康展, 千葉豪, 久語輝彦, 石川眞, 日本原子力学会春の年会予稿集(CD-ROM), 2010, ROMBUNNO.G30, 09 Mar. 2010
Japanese

「JENDL‐4の完成と今後の展望」核分裂炉に対するJENDL‐4.0の積分テスト
千葉豪, 奥村啓介, 杉野和輝, 日本原子力学会春の年会予稿集(CD-ROM), 2010, ROMBUNNO.JS03, 09 Mar. 2010
Japanese

Incorporation of Two-Term Rational Approximation in Tone Method for Resonance Calculation
YAMAMOTO Akio, ENDO Tomohiro, CHIBA Go, Trans Am Nucl Soc, 103, 711, 713, 2010
English

Towards the Completion of JENDL-4: Current Status of Integral Test
CHIBA Go, OKUMURA Keisuke, 日本原子力研究開発機構JAEA-Conf, 21, 26, Oct. 2009
English

Benchmarking of Effective Delayed Neutron Fraction With Deterministic Method
IRWANTO Dwi, OBARA Toru, CHIBA Go, 日本原子力研究開発機構JAEA-Conf, 151, 156, Oct. 2009
English

Nuclear Data-Induced Uncertainty Calculation for Fast Reactor Eigenvalue Separation
CHIBA Go, 日本原子力研究開発機構JAEA-Conf, 101, 106, Oct. 2009
English

Criticality Calculations with Fission Spectrum Matrix
CHIBA Go, 日本原子力研究開発機構JAEA-Conf, 107, 110, Oct. 2009
English

軽水炉使用済燃料核種インベントリに対する燃焼感度解析
千葉豪, 奥村啓介, 日本原子力学会秋の大会予稿集(CD-ROM), 2009, F26, 28 Aug. 2009
Japanese

Calculations of Kobayashi Benchmark Problems with a Neutron Transport Simulation Code System CBG
CHIBA GO, 日本原子力研究開発機構JAEA-Research, 2009, 12, 51P, 43,巻頭1〜2, Jul. 2009
日本原子力研究開発機構, Japanese

Calculations of Kobayashi benchmark problems with a neutron transport simulation code system CBG
千葉 豪, JAEA-research, 2009, 12, 1, 43,巻頭1〜2, Jul. 2009
日本原子力研究開発機構, Japanese

Study on Calculation Methods for Effective Delayed Neutron Fraction
IRWANTO Dwi, OBARA Toru, CHIBA Go, NAGAYA Yasunobu, 日本原子力学会春の年会予稿集(CD-ROM), 2009, K30, 06 Mar. 2009
English

原子炉計算用多群ライブラリの精度向上方策の検討(2)
千葉豪, 奥村啓介, 日本原子力学会春の年会予稿集(CD-ROM), 2009, K45, 06 Mar. 2009
Japanese

Convariances of Resonance Self-Shielding Factor and its Temperature Gradient for Uncertainty Evaluation of Doppler Reactivity
ZUKERAN Atsushi, CHIBA Go, OTUKA Naohiko, ISHIKAWA Makoto, TAKANO Hideki, 日本原子力研究開発機構JAEA-Research, 2008, 91, 174P, 162,巻頭1〜2, Feb. 2009
日本原子力研究開発機構, English

Covariances of resonance self-shielding factor and its temperature gradient for uncertainty evaluation of Doppler reactivity
瑞慶覧 篤, 千葉 豪, 大塚 直彦, JAEA-research, 2008, 91, 1, 162,巻頭1〜2, Feb. 2009
日本原子力研究開発機構, English

Covariance Evaluation of Self-Shielding Factor and Its Temperature Gradient for Uncertainty Evaluation of Doppler Reactivity
OTSUKA Naohiko, OTSUKA Naohiko, ZUKERAN Atsushi, TAKANO Hideki, CHIBA Go, ISHIKAWA Makoto, 日本原子力研究開発機構JAEA-Conf, 32, 40, Nov. 2008
English

CBGLIB: A Multi-group Neutron Library for Precise Neutronics Simulations
CHIBA Go, 日本原子力研究開発機構JAEA-Conf, 114, 120, Nov. 2008
English

Nuclear Data Benchmark for Sodium Voided Reactivity Worth with Improved Neutronics Simulation Method
CHIBA Go, 日本原子力研究開発機構JAEA-Conf, 108, 113, Nov. 2008
English

Benchmark Test of Evaluated Nuclear Data Files for Development of JENDL-4
CHIBA GO, OKUMURA KEISUKE, 日本原子力研究開発機構JAEA-Research, 2008, 89, 58P, 48,巻頭1〜9, Nov. 2008
日本原子力研究開発機構, Japanese

Benchmark test of evaluated nuclear data files for development of JENDL-4
千葉 豪, 奥村 啓介, JAEA-research, 2008, 89, 1, 48,巻頭1〜9, Nov. 2008
日本原子力研究開発機構, Japanese

Nuclear Data Needs for Fast Reactors
CHIBA Go, 日本原子力研究開発機構JAEA-Conf, 26, 31, Sep. 2008
English

核分裂スペクトルに関する実効増倍率の不確定性についての考察
長家康展, 千葉豪, 石川眞, 日本原子力学会秋の大会予稿集(CD-ROM), 2008, A23, 21 Aug. 2008
Japanese

ERRORF-A Code to Calculate Covariance of Self-shielding Factor and Its Temperature Gradient
OTUKA Naohiko, ZUKERAN Atsushi, TAKANO Hideki, CHIBA Go, ISHIKAWA Makoto, 日本原子力研究開発機構JAEA-Data/Code, 21P, Jun. 2008
English

モジュラー型炉心解析コードシステムMOSRAの開発
奥村啓介, 久語輝彦, 中野佳洋, 長家康展, 小嶋健介, 千葉豪, 岡嶋成晃, 日本原子力学会春の年会予稿集(CD-ROM), 2008, L46, 11 Mar. 2008
Japanese

原子炉計算用多群ライブラリの精度向上方策の検討
千葉豪, 奥村啓介, 日本原子力学会春の年会予稿集(CD-ROM), 2008, L30, 11 Mar. 2008
Japanese

積分データによる理論計算パラメータ及び微分核データの調整
千葉敏, 岩本修, 中川庸夫, 千葉豪, 奥村啓介, 日本原子力学会春の年会予稿集(CD-ROM), 2008, B20, 11 Mar. 2008
Japanese

中性子輸送ベンチマーク問題のための参照解の計算
千葉豪, 沼田一幸, 日本原子力研究開発機構JAEA-Review, 119, Feb. 2008
Japanese

マイナーアクチニド核データの積分テスト
千葉豪, 奥村啓介, 日本原子力学会秋の大会予稿集(CD-ROM), 2007, M12, 07 Sep. 2007
Japanese

Benchmark Test of Evaluated Nuclear Data Files for Fast Reactor Neutronics Application
CHIBA GO, IWAI TAKEHIKO, NUMATA KAZUYUKI, HAZAMA TAIRA, 日本原子力研究開発機構JAEA-Research, 62P, Jul. 2007
Japanese

ロスアラモス体験記-核データ部会「若手研究者LANL夏期国際交流プログラム」報告
千葉 豪, 日本原子力学会誌 = Journal of the Atomic Energy Society of Japan, 49, 3, 211, 212, 28 Mar. 2007
Japanese

分離及び非分離共鳴領域における共分散行列の評価と処理(SG‐20)
千葉豪, 日本原子力学会春の年会要旨集(CD-ROM), 2007, TD02, 06 Mar. 2007
Japanese

高速炉用オブジェクト統合型解析システムの研究開発―(7)中性子輸送ソルバーの開発と検証―
千葉豪, 日本原子力学会春の年会要旨集(CD-ROM), 2007, J35, 06 Mar. 2007
Japanese

反射体の材質変更による「常陽」運転用燃料の利用効率改善
関根隆, 大川内靖, 千葉豪, 伊東秀明, 青山卓史, 日本原子力学会春の年会要旨集(CD-ROM), 2007, H38, 06 Mar. 2007
Japanese

ERRORJ: A Code to Process Neutron-nuclide Reaction Cross Section Covariance, Version 2.3
CHIBA Go, 日本原子力研究開発機構JAEA-Data/Code, 26P, Mar. 2007
English

鉄に感度を有する高速炉炉心の臨界性解析
千葉豪, 日本原子力学会和文論文誌, 4, 1, 66, 76, 25 Mar. 2005
日本原子力学会, Japanese

高速実験炉「常陽」MK‐III性能試験‐制御棒校正(NT‐321)‐
千葉豪, 北野彰洋, 前田茂貴, 関根隆, 核燃料サイクル開発機構公開資料, 90P, Oct. 2004
Japanese

工学系モデリング言語としての次世代解析システムの開発(IV)‐境界要素法に基づく中性子拡散ソルバーの開発‐
千葉豪, 核燃料サイクル開発機構公開資料, 54P, Oct. 2004
Japanese

工学系モデリング言語としての次世代解析システムの開発(III)‐プロトタイプ作成による検討(その2)‐
横山賢治, 細貝広視, 千葉豪, 笠原直人, 石川真, 核燃料サイクル開発機構公開資料, 164P, Apr. 2004
Japanese

「常陽」MK‐III性能試験―(4)制御棒反応度価値―
千葉豪, 加藤優子, 服部和裕, 関根隆, 日本原子力学会春の年会要旨集, 42nd, 525, 11 Mar. 2004
Japanese

SLAROM‐UF:高速炉用超微細群格子計算コード(マニュアル)
羽様平, 千葉豪, 佐藤若英, 沼田一幸, 核燃料サイクル開発機構公開資料, 94P, Mar. 2004
Japanese

"JOYO" MK-III Core Performance Tests and Results
MAEDA SHIGETAKA, YOKOYAMA KENJI, CHIBA GO, SEKINE TAKASHI, AOYAMA TAKAFUMI, サイクル機構技報, 21, 99, 109, 20 Dec. 2003
Japanese

MK-3性能試験計画と試験結果 (特集 高速実験炉「常陽」の高度化計画(MK-3計画)と今後の展望) -- (総合機能試験と性能試験)
前田 茂貴, 横山 賢治, 千葉 豪, サイクル機構技報, 0, 21, 99, 109, Dec. 2003
核燃料サイクル開発機構技術展開部, Japanese

サブグループパラメータ計算コードMOMENTOFの開発
千葉豪, 核燃料サイクル開発機構公開資料, 27P, Jul. 2003
Japanese

高速炉体系におけるJENDL‐3.3の積分テスト
千葉豪, 羽様平, 佐藤若英, 沼田一幸, 核燃料サイクル開発機構公開資料, 48P, May 2003
Japanese

高速炉核特性に対するJENDL‐3.3の積分テスト
羽様平, 千葉豪, 日本原子力学会春の年会要旨集, 41st, 405, 05 Mar. 2003
Japanese

高速炉における格子計算手法の比較
千葉豪, 日本原子力学会春の年会要旨集, 41st, 395, 05 Mar. 2003
Japanese

Computer Code System for the R&D of Nuclear Fuel Cycle with Fast Reactor. V. Development and Application of Reactor Analysis Code System.
YOKOYAMA KENJI, HAZAMA TAIRA, CHIBA GO, OKI SHIGEO, ISHIKAWA MAKOTO, サイクル機構技報, 17, 19, 30, 20 Dec. 2002
核燃料サイクル開発機構技術展開部, Japanese

高速炉核特性解析における炉定数高度化の効果に関する検討
千葉豪, 核燃料サイクル開発機構公開資料, 30P, Dec. 2002
Japanese

技術報告 高速炉サイクルの研究開発を支える解析システム(5)核特性解析コードシステムの開発と利用
横山 賢治, 羽様 平, 千葉 豪, サイクル機構技報, 0, 17, 19, 30, Dec. 2002
核燃料サイクル開発機構技術展開部, Japanese

統合炉定数ADJ2000の使用に係る炉定数効果の評価(研究報告書)
庄野彰, 千葉豪, 沼田一幸, 核燃料サイクル開発機構公開資料, 38P, Nov. 2002
Japanese

高速炉用統合炉定数ADJ2000Rの作成 (研究報告)
羽様平, 千葉豪, 沼田一幸, 佐藤若英, 核燃料サイクル開発機構公開資料, 333P, Nov. 2002
Japanese

高速炉格子計算手法の比較研究
千葉豪, 核燃料サイクル開発機構公開資料, 87P, Oct. 2002
Japanese

次世代炉定数システムによる高速炉核特性解析 (II) 次世代炉定数システム導入による効果
千葉豪, 杉野和輝, 石川真, 沼田一幸, 日本原子力学会春の年会要旨集, 40th, 285, 05 Mar. 2002
Japanese

次世代炉定数システムによる高速炉核特性解析 (I) JFS‐3‐J3.2作成時の重み関数訂正効果
千葉豪, 杉野和輝, 石川真, 沼田一幸, 日本原子力学会春の年会要旨集, 40th, 284, 05 Mar. 2002
Japanese

JENDL‐3.2に基づく高速炉用炉定数JFS‐3‐J3.2Rの作成
千葉豪, 沼田一幸, 核燃料サイクル開発機構公開資料, 45P, Feb. 2002
Japanese

The revision of nuclear constant set for fast reactor, JFS-3-J3.2
Gou Chiba, Taira Hazama, Makoto Ishikawa, Transactions of the Atomic Energy Society of Japan, 1, 4, 335, 340, 2002
The fast reactor group constant set JFS-3-J3.2 based on the evaluated nuclear data library JENDL-3.2 has been widely used in fast reactor analysis. However, it was recently found that there were errors in the process of making the group constant and they were revised. This set is called JFS-3-J3.2R. In this report, effects of the errors on nuclear characteristics were evaluated by a comparison with a new reactor group constant set, JFS-3-J3.2R. This report shows that the errors mainly affect removal cross section and distort neutron spectrum. As a result nuclear characteristics, such as sample Doppler reactivity and reaction rate in a blanket region, are significantly affected. However, it is also shown that other characteristics, such as criticality and sodium void reactivity, are not affected because the effects of errors are canceled out as a total integrated result. © 2002, Atomic Energy Society of Japan. All rights reserved., 日本原子力学会, English

Evaluation of an Effect on Nuclear Characteristics by Correcting the Weighting Function in JFS-3-J3.2.
CHIBA GO, NUMATA KAZUYUKI, JFS-3-J3.2の重み関数訂正による核特性への影響評価 (研究報告) 平成13年, 60P, 2001
Japanese

階層領域分割型境界要素法による中性子拡散方程式解法の並列化 (II) 通信の並列化による高速化
辻雅司, 千葉豪, 日本原子力学会春の年会要旨集, 37th, 386, 22 Mar. 1999
Japanese

原子炉の物理　　　　　　　　　　　　　　　
11章（発熱と伝熱、発電）
日本原子力学会, Dec. 2019, [Contributor]

原子力発電プラント構造物の放射化計算のための革新的超高速計算フレームワーク
科学研究費助成事業
01 Apr. 2022 - 31 Mar. 2025
千葉 豪
日本学術振興会, 基盤研究(C), 北海道大学, 22K04979

TM-EFP: an inovative numerical model for nuclear reactor transient analyses
Grants-in-Aid for Scientific Research
01 Apr. 2016 - 31 Mar. 2019
Chiba Go
Delayed neutrons play an important role in nuclear energy utilization by human being, and delayed neutron emission is important physical phenomena which should be considered especially in safety analyses of nuclear reactors. The conventional model which has been widely and generally used in the world introduces an approximation that various fission product nuclides are treated as small number of fictitious nuclides, but we propose and develop a more sophisticated model which can handle with all the fission product nuclides explicitly and realize spatially-dependent nuclear reactor kinetics calculations with this model. As an example, nuclear reactor transient problems with leakage of gaseous fission product nuclides are carried out, and effectiveness of this model is demonstrated.
Japan Society for the Promotion of Science, Grant-in-Aid for Young Scientists (B), Hokkaido University, 16K18344

Uncertainty quantification of reactor analysis method at design extension condition: A new estimation method based on covariance between input and prediction error
Grants-in-Aid for Scientific Research
01 Apr. 2016 - 31 Mar. 2019
Yamamoto Akio, Khalik Abdel Hany
A new evaluation method for prediction error of neutronics simulations under severe accident conditions of a nuclear reactor has been developed. The present method utilizes the Kriging method, which is used in the field of geostatstics, and the principal component analysis. Calculation error of simulation is evaluated by learning the correlation between the parameters used in the simulation and the calculation errors.
The present method is applied to predict the error of effective multiplication factor of LWR fuel assemblies for various conditions including from normal operation to severe accident conditions. The difference of multiplication factors obtained by the deterministic and the continuous energy Monte-Carlo methods is considered as the calculation error. The results indicate that the present method accurately estimates the calculation error for wide range of reactor conditions.
Japan Society for the Promotion of Science, Grant-in-Aid for Scientific Research (C), Nagoya University, 16K06956

Uncertainty estimation of unmeasurable core safety parameters - a new approach using covariance matrix
Grants-in-Aid for Scientific Research
01 Apr. 2012 - 31 Mar. 2015
YAMAMOTO Akio, CHIBA Go, ENDO Tomohiro
In the present study, uncertainty estimation of unmeasurable core neutronics parameters is studied, especially theoretical framework and implementation of numerical methods, and validity of the developed methods is confirmed through verifications. A new approach for the cross section adjustment and the bias factor method is developed based on the random sampling method in order to apply them to actual light water reactors. All goals of the present study are fulfilled and the above achievements are beyond the plan at the beginning of this study.
Japan Society for the Promotion of Science, Grant-in-Aid for Scientific Research (C), Nagoya University, 24561040

Development of High Efficiency Filtered Containment Venting System by using AgX
Grants-in-Aid for Scientific Research
01 Apr. 2012 - 31 Mar. 2015
NARABAYASHI Tadashi, SATO Masanobu, TSUJI Masashi, CHIBA Go
From the lessons of TMI and Chernobyl Accidents, filtered containment venting system (FCVS) and Containment vessel spray cooling system are enforced in the the New Regulatory Requirement, by using mobile generators and heat exchangers to keep the ultimate heat sink even in any natural disaster, such as large earthquake, big tsunami etc. In this study we have succeeded to develop high decontamination factor FCVS that used Silver Zeolite named AgX. Hokkaido University has tested wet type FCVS using venturi scrubber in hot water pool and dry type FCVS using metallic fiber filter for 1st stage and AgX for 2nd stage, generating super heat steam by using multi-stage orifice. Tohoku University succeeded to use silver doped zeolite (AgX) which is a promising sorbent for CH3I can remove over 99.99%. The results are the 1st class high decontamination factor and applied to actual Nuclear power plants in Japan.
Japan Society for the Promotion of Science, Grant-in-Aid for Scientific Research (B), Hokkaido University, 24360388