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

  • 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]
  • 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]
  • Chiba Go
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 55 3 290 - 300 2018年 [査読有り][通常論文]
  • Endo Tomohiro, Chiba Go, van Rooijen Willem Frederik Geert, Yamanaka Masao, Pyeon Cheol Ho
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 55 4 450 - 459 2018年 [査読有り][通常論文]
  • Miwa Shuichiro, Yamamoto Yasunori, Chiba Go
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 55 6 575 - 598 2018年 [査読有り][通常論文]
  • Pyeon Cheol Ho, Yamanaka Masao, Ito Makoto, Chiba Go, Endo Tomohiro, Kim Song Hyun, van Rooijen Willem Fredrik G
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 55 7 812 - 821 2018年 [査読有り][通常論文]
  • Chiba Go, Okumura Shintaro
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 55 9 1043 - 1053 2018年 [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.
  • 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.
  • 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.
  • Chiba Go, Narabayashi Tadashi
    ANNALS OF NUCLEAR ENERGY 85 846 - 855 2015年11月 [査読有り][通常論文]
  • Chiba Go, Tsuji Masashi, Narabayashi Tadashi, Ohoka Yasunori, Ushio Tadashi
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 52 7-8 953 - 960 2015年08月03日 [査読有り][通常論文]
  • Kawamoto Yosuke, Chiba Go, Tsuji Masashi, Narabayashi Tadashi
    ANNALS OF NUCLEAR ENERGY 80 219 - 224 2015年06月 [査読有り][通常論文]
  • Chiba Go, Kawamoto Yosuke, Tsuji Masashi, Narabayashi Tadashi
    ANNALS OF NUCLEAR ENERGY 75 395 - 403 2015年01月 [査読有り][通常論文]
  • 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.
  • CHIBA Go, TSUJI Masashi, NARABAYASHI Tadashi
    J Nucl Sci Technol 50 11-12 1150 - 1160 2013年11月 [査読無し][通常論文]
  • CHIBA Go, TSUJI Masashi, NARABAYASHI Tadashi
    J Nucl Sci Technol 50 7-8 751 - 760 2013年07月 [査読無し][通常論文]
  • 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.
  • CHIBA Go, TSUJI Masashi, SUGIYAMA Ken-ichiro, NARABAYASHI Tadashi
    J Nucl Sci Technol 48 12 1471 - 1477 Atomic Energy Society of Japan 2011年12月01日 [査読無し][通常論文]
  • 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]
  • 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.
  • 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.
  • CHIBA Go, NAGAYA Yasunobu, MORI Takamasa
    J Nucl Sci Technol 48 8 1163 - 1169 2011年 [査読無し][通常論文]
  • 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.
  • SHIBATA Keiichi, CHIBA Go, ICHIHARA Akira, KUNIEDA Satoshi
    J Nucl Sci Technol 47 1 40 - 46 2010年 [査読無し][通常論文]
  • CHIBA Go, OKUMURA Keisuke, OIZUMI Akito, SAITO Masaki
    J Nucl Sci Technol 47 7 652 - 660 2010年 [査読無し][通常論文]
  • 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.
  • Chiba Go, Nagaya Yasunobu
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 46 10 1000 - 1003 2009年10月 [査読有り][通常論文]
  • NAGAYA Yasunobu, KODELI Ivan, CHIBA Go, ISHIKAWA Makoto
    Nucl Instrum Methods Phys Res Sect A Accel Spectrometers Detect Assoc 603 3 485 - 490 2009年05月21日 [査読有り][通常論文]
  • Chiba Go
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 46 5 399 - 402 2009年05月 [査読有り][通常論文]
  • Iwamoto Osamu, Nakagawa Tsuneo, Otuka Naohiko, Chiba Satoshi, Okumura Keisuke, Chiba Go, Ohsawa Takaaki, Furutaka Kazuyoshi
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 46 5 510 - 528 2009年05月 [査読有り][通常論文]
  • IWAMOTO Osamu, NAKAGAWA Tsuneo, OTUKA Naohiko, CHIBA Satoshi, OKUMURA Keisuke, CHIBA Go, OHSAWA Takaaki, FURUTAKA Kazuyoshi
    J Nucl Sci Technol 46 5 510 - 528 2009年 [査読無し][通常論文]
  • CHIBA Go, NAGAYA Yasunobu
    J Nucl Sci Technol 46 10 1000 - 1003 2009年 [査読無し][通常論文]
  • CHIBA Go
    J Nucl Sci Technol 46 5 399 - 402 2009年 [査読無し][通常論文]
  • OTUKA Naohiko, OTUKA Naohiko, ZUKERAN Atsushi, TAKANO Hideki, CHIBA Go, CHIBA Go, ISHIKAWA Makoto
    J Nucl Sci Technol 45 3 195 - 210 2008年03月 [査読有り][通常論文]
  • 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.
  • 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03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 3 2 200 - 207 2004年06月25日 [査読無し][通常論文]
  • 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12月25日 [査読無し][通常論文]

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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