研究者データベース

千葉 豪(チバ ゴウ)
工学研究院 応用量子科学部門 量子エネルギー工学
准教授

基本情報

所属

  • 工学研究院 応用量子科学部門 量子エネルギー工学

職名

  • 准教授

学位

  • 工学博士(北海道大学)

J-Global ID

研究分野

  • エネルギー / 原子力工学

職歴

  • 2014年04月 - 現在 北海道大学 工学研究院 准教授
  • 2011年04月 - 2014年03月 北海道大学 工学研究院 助教
  • 2005年10月 - 2011年03月 日本原子力研究開発機構 研究員
  • 2001年04月 - 2005年09月 核燃料サイクル開発機構 研究員

学歴

  • 1999年04月 - 2001年03月   北海道大学   大学院工学研究科
  • 1995年04月 - 1999年03月   北海道大学   工学部

研究活動情報

論文

  • Cheol Ho Pyeon, Masao Yamanaka, Tomohiro Endo, Go Chiba, Willem F. G Van Rooijen, Kenichi Watanabe
    Nuclear Science and Engineering 1 - 12 2020年07月07日 [査読有り][通常論文]
  • Katagiri Koji, Chiba Go
    ANNALS OF NUCLEAR ENERGY 133 202 - 208 2019年11月 [査読有り][通常論文]
  • Yamanaka Masao, Pyeon Cheol Ho, Endo Tomohiro, Watanabe Kenichi, Chiba Go, van Rooijen Willem Frederik Geert
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2019年09月26日 [査読有り][通常論文]
  • 藤井 宏之, 千葉 豪, 山田 幸生, 星 詳子, 小林 一道, 渡部 正夫
    Proceedings of the 9th International Symposium on Radiative Transfer, RAD-19 SM05 261 - 268 2019年09月 [査読有り][通常論文]
  • Endo Tomohiro, Watanabe Kenichi, Chiba Go, Yamanaka Masao, van Rooijen Willem Frederik Geert, Pyeon Cheol Ho
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2019年08月15日 [査読有り][通常論文]
  • Chiba Go, Endo Tomohiro
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2019年08月05日 [査読有り][通常論文]
  • 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 2019年08月03日 [査読有り][通常論文]
  • Nihira Shunsuke, Chiba Go
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2019年06月19日 [査読有り][通常論文]
  • Array,Yukio Yamada, Go Chiba, Yoko Hoshi, Array, Masao Watanabe
    J. Comput. Physics 374 591 - 604 2018年12月01日 [査読有り][通常論文]
  • Chiba Go, Nihira Shunsuke
    EPJ NUCLEAR SCIENCES & TECHNOLOGIES 4 2018年11月14日 [査読有り][通常論文]
  • 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 2018年07月03日 [査読有り][通常論文]
     
    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.
  • Shuichiro Miwa, Yasunori Yamamoto, Go Chiba
    Journal of Nuclear Science and Technology 55 6 575 - 598 2018年06月03日 [査読有り][通常論文]
     
    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.
  • Tomohiro Endo, Go Chiba, Willem Frederik Geert van Rooijen, Masao Yamanaka, Cheol Ho Pyeon
    Journal of Nuclear Science and Technology 55 4 450 - 459 2018年04月03日 [査読有り][通常論文]
     
    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.
  • Go Chiba
    Journal of Nuclear Science and Technology 55 3 290 - 300 2018年03月04日 [査読有り][通常論文]
     
    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.
  • Chiba Go, Okumura Shintaro
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 55 9 1043 - 1053 2018年 [査読有り][通常論文]
  • Shotaro Imai, Ankit Ravankar, Michiyo Shimamura, Taichi Takasuka, Go Chiba, Yasuhiro Yamanaka
    International Journal of Institutional Research and Management 1 2 17 - 35 2017年09月 [査読有り][通常論文]
  • W. F. G. Rooijen, T. Endo, G. Chiba, C. H. Pyeon
    Progress in Nuclear Energy 101 243 - 250 2017年09月 [査読有り][通常論文]
  • 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 2017年09月 [査読有り][通常論文]
     
    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.
  • Cheol Ho Pyeon, Masao Yamanaka, Tomohiro Endo, Willem Fredrik G. van Rooijen, Go Chiba
    ANNALS OF NUCLEAR ENERGY 105 346 - 354 2017年07月 [査読有り][通常論文]
     
    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.
  • Ankit A. Ravankar, Shotaro Imai, Michiyo Shimamura, Go Chiba, Taichi Takasuka
    高等教育ジャーナル-高等教育と生涯学習- 24 9 - 20 2017年03月 [査読有り][通常論文]
  • Go Chiba
    ANNALS OF NUCLEAR ENERGY 101 23 - 30 2017年03月 [査読有り][通常論文]
     
    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.
  • Yosuke Kawamoto, Go Chiba
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 54 2 213 - 222 2017年 [査読有り][通常論文]
     
    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.
  • Go Chiba, Yosuke Kawamoto, Tadashi Narabayashi
    ANNALS OF NUCLEAR ENERGY 96 313 - 323 2016年10月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Tadashi Narabayashi
    ANNALS OF NUCLEAR ENERGY 96 277 - 286 2016年10月 [査読有り][通常論文]
     
    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.
  • Takanori Kajihara, Masashi Tsuji, Go Chiba, Yosuke Kawamoto, Yasunori Ohoka, Tadashi Ushio
    ANNALS OF NUCLEAR ENERGY 94 742 - 749 2016年08月 [査読有り][通常論文]
     
    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.
  • 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 2016年 [査読有り][通常論文]
     
    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.
  • Go Chiba, Cheol Ho Pyeon, Wilfred van Rooijen, Tomohiro Endo
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 53 10 1653 - 1661 2016年 [査読有り][通常論文]
     
    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.
  • Go Chiba, Tadashi Narabayashi
    ANNALS OF NUCLEAR ENERGY 85 846 - 855 2015年11月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Masashi Tsuji, Tadashi Narabayashi, Yasunori Ohoka, Tadashi Ushio
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 52 7-8 953 - 960 2015年08月 [査読有り][通常論文]
     
    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.
  • Yosuke Kawamoto, Go Chiba, Masashi Tsuji, Tadashi Narabayashi
    ANNALS OF NUCLEAR ENERGY 80 219 - 224 2015年06月 [査読有り][通常論文]
     
    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.
  • Masahide Kobayashi, Tadashi Narabayashi, Masashi Tuji, Go Chiba, Yasunori Nagata, Tomohiro Shimoe
    Transactions of the Atomic Energy Society of Japan 14 1 12 - 24 2015年 [査読有り][通常論文]
     
    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.
  • G. Chiba, M. Tsuji, T. Narabayashi
    NUCLEAR DATA SHEETS 123 62 - 67 2015年01月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Yosuke Kawamoto, Masashi Tsuji, Tadashi Narabayashi
    ANNALS OF NUCLEAR ENERGY 75 395 - 403 2015年01月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Masashi Tsuji, Tadashi Narabayashi
    NUCLEAR ENGINEERING AND TECHNOLOGY 46 3 281 - 290 2014年06月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Masashi Tsuji, Tadashi Narabayashi
    ANNALS OF NUCLEAR ENERGY 65 41 - 46 2014年03月 [査読無し][通常論文]
     
    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.
  • Masahiro Tatsumi, Go Chiba
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 51 9 1161 - 1163 2014年 [査読有り][通常論文]
     
    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.
  • Go Chiba, Masashi Tsuji, Tadashi Narabayashi
    Journal of Nuclear Science and Technology 50 12 1150 - 1160 2013年12月01日 [査読無し][通常論文]
     
    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.
  • Go Chiba, Masashi Tsuji, Tadashi Narabayashi
    Journal of Nuclear Science and Technology 50 7 751 - 760 2013年07月01日 [査読無し][通常論文]
     
    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.
  • Go Chiba, Akio Yamamoto, Masashi Tsuji, Tadashi Narabayashi
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 49 7 748 - 753 2012年07月 [査読無し][通常論文]
     
    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.
  • Go Chiba, Masashi Tsuji, Ken-ichiro Sugiyama, Tadashi Narabayashi
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 49 1-2 272 - 280 2012年01月 [査読無し][通常論文]
     
    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.
  • Go Chiba, Masashi Tsuji, Ken-ichiro Sugiyama, Tadashi Narabayashi
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 48 12 1471 - 1477 2011年12月 [査読無し][通常論文]
     
    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.
  • Go Chiba, Masashi Tsuji, Ken-ichiro Sugiyama, Tadashi Narabayashi
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 48 12 1471 - 1477 2011年12月 [査読有り][通常論文]
     
    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.
  • 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  2011年10月 [査読無し][通常論文]
  • Go Chiba, Yasunobu Nagaya, Takamasa Mori
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 48 8 1163 - 1169 2011年08月 [査読無し][通常論文]
     
    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.
  • 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 2011年08月 [査読有り][通常論文]
     
    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.
  • 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 2011年08月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Yasunobu Nagaya, Takamasa Mori
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 48 8 1163 - 1169 2011年08月 [査読有り][通常論文]
     
    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.
  • Go Chiba
    ANNALS OF NUCLEAR ENERGY 38 5 1033 - 1038 2011年05月 [査読有り][通常論文]
     
    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.
  • 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 2011年02月 [査読有り][通常論文]
     
    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.
  • Akio Yamamoto, Tomohiro Endo, Go Chiba
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 48 2 263 - 271 2011年02月 [査読有り][通常論文]
     
    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.
  • W. F. G. van Rooijen, G. Chiba
    ANNALS OF NUCLEAR ENERGY 38 1 133 - 144 2011年01月 [査読有り][通常論文]
     
    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
  • Yasunobu Nagaya, Go Chiba, Takamasa Mori, Dwi Irwanto, Ken Nakajima
    ANNALS OF NUCLEAR ENERGY 37 10 1308 - 1315 2010年10月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Keisuke Okumura, Akito Oizumi, Masaki Saito
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 47 7 652 - 660 2010年07月 [査読無し][通常論文]
     
    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.
  • Go Chiba, Keisuke Okumura, Akito Oizumi, Masaki Saito
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 47 7 652 - 660 2010年07月 [査読有り][通常論文]
     
    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.
  • Keiichi Shibata, Go Chiba, Akira Ichihara, Satoshi Kunieda
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 47 1 40 - 46 2010年01月 [査読無し][通常論文]
     
    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.
  • Keiichi Shibata, Go Chiba, Akira Ichihara, Satoshi Kunieda
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 47 1 40 - 46 2010年01月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Yasunobu Nagaya
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 46 10 1000 - 1003 2009年10月 [査読無し][通常論文]
     
    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.
  • Go Chiba, Yasunobu Nagaya
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 46 10 1000 - 1003 2009年10月 [査読有り][通常論文]
     
    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.
  • 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 2009年05月 [査読無し][通常論文]
     
    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.
  • 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 2009年05月 [査読有り][通常論文]
     
    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.
  • Go Chiba
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 46 5 399 - 402 2009年05月 [査読有り][通常論文]
  • 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 2009年05月 [査読有り][通常論文]
     
    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.
  • Go Chiba
    Journal of Nuclear Science and Technology 46 5 399 - 402 2009年 [査読無し][通常論文]
     
    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.
  • 岩本 修, 中川 庸雄, 大塚 直彦, 千葉 敏, 奥村 啓介, 千葉 豪
    Nuclear Data Sheets 109 12 2885 - 2889 2008年12月 [査読有り][通常論文]
     
    JENDLアクチノイドファイル2008(JENDL/AC-2008)を2008年3月に公開した。ファイルにはAcからFmまでのアクチノイド核種に対する中性子誘起核反応データが含まれている。JENDL-3.3の62核種のデータを改訂するとともに、半減期1日以上の17核種を新たに追加した。高速エネルギー領域の断面積や2次中性子のエネルギー・角度分布の評価に、核反応モデルコードCCONEを広く使用した。JENDL/AC-2008の重要核種に対する核分裂断面積,捕獲断面積,核分裂中性子数の共分散を評価する予定である。評価手法及び結果また共分散評価計画について発表を行う。
  • JENDLアクチノイドファイルの開発
    岩本 修, 中川 庸雄, 大塚 直彦, 千葉 敏, 奥村 啓介, 千葉 豪
    Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM) 8  2008年09月 [査読無し][通常論文]
     
    JENDLアクチノイドファイル2008(JENDL/AC 2008)のためAcからFmのアクチニドに対する中性子誘起反応の核データを評価した。入手可能な測定データ及び新しく開発した理論モデルコードCCONEを用いて評価を行い、JENDL-3.3のほぼすべてのデータを更新した。JENDL/ACの予備的なバージョンを使用した核分裂炉に対する積分ベンチマークテストを行っている。JENDL/ACは2008年に公開予定である。
  • Naohiko Otuka, Atsushi Zukeran, Hideki Takano, Go Chiba, Makoto Ishikawa
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 45 3 195 - 210 2008年03月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Yoichiro Shimazu
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 44 12 1526 - 1534 2007年12月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Kazuyuki Numata
    ANNALS OF NUCLEAR ENERGY 34 6 443 - 448 2007年06月 [査読有り][通常論文]
     
    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.
  • Go Chiba
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 43 11 1395 - 1405 2006年11月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Hironobu Unesaki
    ANNALS OF NUCLEAR ENERGY 33 13 1141 - 1146 2006年09月 [査読有り][通常論文]
     
    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.
  • Go Chiba
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 43 8 946 - 949 2006年08月 [査読有り][通常論文]
  • Taira Hazama, Go Chiba, Kazuteru Sugino
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 43 8 908 - 918 2006年08月 [査読有り][通常論文]
     
    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.
  • Go Chiba, Makoto Ishikawa
    AIP Conference Proceedings 769 468 - 471 2005年05月24日 [査読有り][通常論文]
     
    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.
  • 千葉 豪
    日本原子力学会和文論文誌 = Transactions of the Atomic Energy Society of Japan 4 1 66 - 76 日本原子力学会 2005年 [査読無し][通常論文]
     
    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.
  • 千葉豪
    日本原子力学会和文論文誌 3 2 200 - 207 2004年06月25日 [査読無し][通常論文]
  • 千葉 豪
    日本原子力学会和文論文誌 = Transactions of the Atomic Energy Society of Japan 3 2 200 - 207 日本原子力学会 2004年 [査読無し][通常論文]
     
    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.
  • G Chiba
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 40 7 537 - 543 2003年07月 [査読有り][通常論文]
     
    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.
  • 千葉豪, 羽様平, 石川真
    日本原子力学会和文論文誌 1 4 335 - 340 2002年 [査読無し][通常論文]
     
    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.
  • G Chiba, M Tsuji, Y Shimazu
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 38 8 664 - 673 2001年08月 [査読有り][通常論文]
     
    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.
  • G Chiba, M Tsuji, Y Shimazu
    ANNALS OF NUCLEAR ENERGY 28 9 895 - 912 2001年06月 [査読有り][通常論文]
     
    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.
  • M Tsuji, G Chiba
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 37 5 477 - 485 2000年05月 [査読有り][通常論文]
     
    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.

その他活動・業績

受賞

  • 2012年03月 日本原子力学会 技術賞特賞
     
    受賞者: 千葉豪;岩本修;柴田恵一
  • 2011年09月 日本原子力学会炉物理部会 炉物理部会賞
     
    受賞者: 千葉豪
  • 2006年09月 日本原子力学会核データ部会 核データ部会奨励賞
     
    受賞者: 千葉豪

教育活動情報

主要な担当授業

  • エネルギー工学概論
    開講年度 : 2018年
    課程区分 : 学士課程
    開講学部 : 工学部
  • EES Advanced Laboratory Seminar(エネルギー環境システム工学特別ラボラトリーセミナーE)
    開講年度 : 2018年
    課程区分 : 修士課程
    開講学部 : 工学院
    キーワード : 実験・演習、計測技術、解析技術、流体、熱、燃焼、原子力、エネルギーと社会
  • エネルギー環境システム工学特別ラボラトリーセミナー
    開講年度 : 2018年
    課程区分 : 修士課程
    開講学部 : 工学院
    キーワード : 実験・演習、計測技術、解析技術、流体、熱、燃焼、原子力、エネルギーと社会
  • Nuclear Reactor Engineering(原子炉工学E)
    開講年度 : 2018年
    課程区分 : 修士課程
    開講学部 : 工学院
  • Nuclear Reactor Theory(原子炉物理E)
    開講年度 : 2018年
    課程区分 : 修士課程
    開講学部 : 工学院
    キーワード : 数値計算、中性子、反復解法、固有値方程式、べき乗法
  • 原子炉工学特論
    開講年度 : 2018年
    課程区分 : 修士課程
    開講学部 : 工学院
  • 原子炉特別実験
    開講年度 : 2018年
    課程区分 : 修士課程
    開講学部 : 工学院
    キーワード : 放射線測定、γ線、中性子、臨界、原子炉制御、制御棒価値、反応率分布
  • 原子炉物理特論
    開講年度 : 2018年
    課程区分 : 修士課程
    開講学部 : 工学院
    キーワード : 数値計算、中性子、反復解法、固有値方程式、べき乗法
  • Nuclear and Various Energy Systems(原子力・エネルギーシステムE)
    開講年度 : 2018年
    課程区分 : 修士課程
    開講学部 : 工学院
    キーワード : 原子力、原子力安全、核燃料サイクル、核不拡散、核融合、再生可能エネルギー、次世代自動車
  • 原子力・エネルギーシステム特論
    開講年度 : 2018年
    課程区分 : 修士課程
    開講学部 : 工学院
    キーワード : 原子力、原子力安全、核燃料サイクル、核不拡散、核融合、再生可能エネルギー、次世代自動車
  • EES Advanced Laboratory Seminar(エネルギー環境システム工学特別ラボラトリーセミナーE)
    開講年度 : 2018年
    課程区分 : 博士後期課程
    開講学部 : 工学院
    キーワード : 実験・演習、計測技術、解析技術、流体、熱、燃焼、原子力、エネルギーと社会
  • エネルギー環境システム工学特別ラボラトリーセミナー
    開講年度 : 2018年
    課程区分 : 博士後期課程
    開講学部 : 工学院
    キーワード : 実験・演習、計測技術、解析技術、流体、熱、燃焼、原子力、エネルギーと社会
  • Nuclear Reactor Engineering(原子炉工学E)
    開講年度 : 2018年
    課程区分 : 博士後期課程
    開講学部 : 工学院
  • Nuclear Reactor Theory(原子炉物理E)
    開講年度 : 2018年
    課程区分 : 博士後期課程
    開講学部 : 工学院
    キーワード : 数値計算、中性子、反復解法、固有値方程式、べき乗法
  • 原子炉工学特論
    開講年度 : 2018年
    課程区分 : 博士後期課程
    開講学部 : 工学院
  • 原子炉特別実験
    開講年度 : 2018年
    課程区分 : 博士後期課程
    開講学部 : 工学院
    キーワード : 放射線測定、γ線、中性子、臨界、原子炉制御、制御棒価値、反応率分布
  • 原子炉物理特論
    開講年度 : 2018年
    課程区分 : 博士後期課程
    開講学部 : 工学院
    キーワード : 数値計算、中性子、反復解法、固有値方程式、べき乗法
  • Nuclear and Various Energy Systems(原子力・エネルギーシステムE)
    開講年度 : 2018年
    課程区分 : 博士後期課程
    開講学部 : 工学院
    キーワード : 原子力、原子力安全、核燃料サイクル、核不拡散、核融合、再生可能エネルギー、次世代自動車
  • 原子力・エネルギーシステム特論
    開講年度 : 2018年
    課程区分 : 博士後期課程
    開講学部 : 工学院
    キーワード : 原子力、原子力安全、核燃料サイクル、核不拡散、核融合、再生可能エネルギー、次世代自動車
  • 一般教育演習(フレッシュマンセミナー)
    開講年度 : 2018年
    課程区分 : 学士課程
    開講学部 : 全学教育
    キーワード : 放射線、放射能、放射線被ばく、放射線医療、原子炉、環境放射能、放射性廃棄物処理・処分、廃炉工学、オープン教材
  • 原子炉物理
    開講年度 : 2018年
    課程区分 : 学士課程
    開講学部 : 工学部
  • エネルギー工学概論
    開講年度 : 2018年
    課程区分 : 学士課程
    開講学部 : 工学部
    キーワード : 地球環境、火力発電、原子力発電、再生可能エネルギー、省エネルギー
  • 科学・技術の世界
    開講年度 : 2018年
    課程区分 : 学士課程
    開講学部 : 全学教育
    キーワード : 原子力技術、原子力発電、放射線、医療応用、宇宙探査、材料開発
  • 原子炉工学
    開講年度 : 2018年
    課程区分 : 学士課程
    開講学部 : 工学部
    キーワード : 原子炉設計、核分裂連鎖反応、臨界、熱水力、動特性
  • 機械加工学実習
    開講年度 : 2018年
    課程区分 : 学士課程
    開講学部 : 工学部
    キーワード : 鋳造、溶接、塑性加工、切削加工、研削加工、先端加工


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