Teruhiko KugoProfessor for Academic Advancement

■Researcher basic information

Organization

  • College of Engineering Department of Mechanical Systems Engineering
  • Faculty of Applied Science and Engineering Domain of Mechanical Systems Engineering

Research Areas

  • Energy, Nuclear engineering

Degree

  • 2008年3月　工学博士（大阪大学）
  • 1989年3月　工学修士（大阪大学）

■Research activity information

Paper

  • Benchmark models for criticalities of FCA-IX assemblies with systematically changed neutron spectra
    Fukushima Masahiro; Kitamura Yasunori; Kugo Teruhiko; Okajima Shigeaki, New benchmark models with respect to criticality data are established on the basis of seven uranium-fueled assemblies constructed in the ninth experimental series at the fast critical assembly (FCA) facility. By virtue of these FCA-IX assemblies where the simple combinations of uranium fuel and diluent (graphite and stainless steel) in their core regions were systematically varied, the neutron spectra of these benchmark models cover those of various reactor types, from fast to sub-moderated reactors. The sample calculations of the benchmark models by a continuous-energy Monte Carlo (MC) code showed obvious differences between even the latest versions of two major nuclear data libraries, JENDL-4.0 and ENDF/B-VII.1. The present benchmark models would be well-suited for assessment and improvement of the nuclear data for $^{235}$U, $^{238}$U, graphite, and stainless steel. In addition, the verification of the deterministic method was performed on the benchmark models by comparison with the MC calculations. The present benchmark models are also available to users of deterministic calculation codes for assessment and improvement of nuclear data.
    Journal of Nuclear Science and Technology, Mar. 2016, [Reviewed]
  • Physical mechanism analysis of burnup actinide composition in light water reactor MOX fuel and its application to uncertainty evaluation
    Oizumi Akito; Jin Tomoyuki*; Ishikawa Makoto; Kugo Teruhiko, The uncertainty associated with physical quantities, such as nuclear data, needs to be quantitatively analyzed. The present paper illustrates an analysis methodology to investigate the physical mechanisms of burnup actinide composition with nuclear-data sensitivity based on the generalized depletion perturbation theory. The target in this paper is the MOX fuel of the light water reactor. We start with the discussion of the basic physical mechanisms for burnup actinide compositions using the reaction-rate flow chart on the burnup chain. After that, the physical mechanisms of the productions of $^{244}$Cm and $^{238}$Pu are analyzed in detail with burnup sensitivity calculation. Conclusively, we can identify the source of actinide productions and evaluate the indirect influence of the nuclear reactions if the physical mechanisms of burnup actinide composition are analyzed using the reaction-rate flow chart on the burnup chain and burnup sensitivity calculation. Finally, we demonstrate the usefulness of the burnup sensitivity coefficients in an application to determine the priority of accuracy improvement in nuclear data in combination with the covariance of the nuclear data. In addition, the target actinides and reactions are categorized into patterns according to a sensitivity trend.
    Annals of Nuclear Energy, Jul. 2015, [Reviewed]
  • Options of principles of fuel debris criticality control in Fukushima Daiichi reactors
    Tonoike Kotaro; Sono Hiroki; Umeda Miki; Yamane Yuichi; Kugo Teruhiko; Suyama Kenya, In the Three Mile Island Unit 2 reactor accident, a large amount of fuel debris was formed whose criticality condition is unknown except the possible highest $^{235}$U/U enrichment. The fuel debris had to be cooled and shielded by water in which the minimum critical mass is much smaller than the total mass of fuel debris. To overcome this uncertain situation, the coolant water was borated with sufficient concentration to secure the subcritical condition. The situation is more severe in the damaged reactors of Fukushima Daiichi Nuclear Power Station, where the coolant water flow is practically "once through". Boron must be endlessly added to the water to secure the subcritical condition of the fuel debris, which is not feasible. The water is not borated relying on the circumstantial evidence that the xenon gas monitoring in the containment vessels does not show a sign of criticality. The criticality condition of fuel debris may worsen due to the gradual drop of its temperature, or the change of its geometry by aftershocks or the retrieval work, that may lead the criticality. To avoid criticality and its severe consequences, a certain principle of criticality control must be established. There may be options, such as prevention of the criticality by coolant water boration or by neutronic monitoring, prevention of the severe consequences by intervention measures against criticality, etc. Every option has merits and demerits that must be adequately evaluated toward selection of the best principle.
    Nuclear Back-end and Transmutation Technology for Waste Disposal, Jan. 2015, [Reviewed]
  • Applications of integral benchmark data
    Palmiotti G.*; Briggs J. B.*; Kugo Teruhiko; Trumble E.*; Kahler A. C.*; Lancaster D.*, The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) provide evaluated integral benchmark data that may be used for validation of reactor physics / nuclear criticality safety analytical methods and data, nuclear data testing, advanced modeling and simulation, and safety analysis licensing activities. The handbooks produced by these programs are used in over 30 countries. Five example applications are presented in this paper: (1) Use of IRPhEP Data in Uncertainty Analyses and Cross Section Adjustment, (2) Uncertainty Evaluation Methods for Reactor Core Design at JAEA Using Reactor Physics Experimental Data, (3) Cross Section Data Testing with ICSBEP Benchmarks, (4) Application of Benchmarking Data to a Broad Range of Criticality Safety Problems, and (5) Use of the International Handbook of Evaluated Reactor Physics Benchmark Experiments to Support the Power Industry.
    Nuclear Science and Engineering, Nov. 2014, [Reviewed]
  • Evaluation of large 3600 MWth sodium-cooled fast reactor OECD neutronic benchmarks
    Buiron L.*; Rimpault G*; Fontaine B.*; Kim T. K.*; Stauff N. E.*; Taiwo T. A.*; Yamaji Akifumi*; Gulliford J.*; Fridmann E.*; Pataki I.*; Kereszt\'uri A.*; T\'ota \'A.*; Kugo Teruhiko; Sugino Kazuteru; Uematsu Mari Mariannu; Lin Tan R.*; Kozlowski T.*; Parisi C.*; Ponomarev A.*, Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD, an international collaboration is ongoing on the neutronic analyses of several Generation-IV Sodium-cooled Fast Reactor (SFR) concepts. This paper summarizes the results obtained by participants from institutions of different countries (ANL, CEA, ENEA, HZDR, JAEA, CER, KIT, UIUC) for the large core numerical benchmarks. These results have been obtained using different calculation methods and analysis tools to estimate the core reactivity and isotopic composition evolution, neutronic feedbacks and power distribution. For the different core concepts analyzed, a satisfactory agreement was obtained between participants despite the different calculation schemes used. A good agreement was generally obtained when comparing compositions after burnup, the delayed neutron fraction, the Doppler coefficient, and the sodium void worth. However, some noticeable discrepancies between the k-effective values were observed and are explained in this paper. These are mostly due to the different neutronic libraries employed (JEFF3.1, ENDFB7.0 or JENDL-4.0) and to a lesser extent the calculations methods.
    Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), Sep. 2014, [Reviewed]
  • Evaluation of OECD/NEA/WPRS benchmark on medium size metallic core SRF by deterministic code system; MARBLE and Monte Carlo code: MVP
    Uematsu Mari Mariannu; Kugo Teruhiko; Numata Kazuyuki*, In the frame work of the working party on reactor and system (WPRS) of the OECD/NEA, the benchmark on SFR was conducted. Within the OECD/NEA/WPRS benchmark, study on medium size metallic fuel core was performed using a code system for fast reactor core calculation with deterministic method MARBLE and with a Monte Carlo method MVP. The latest nuclear library JENDL-4.0 is used for evaluation of eigenvalues (k$_{\rm eff}$) and reactivity (sodium void, Doppler and control rod worth) calculations. Depletion calculations are conducted using MARBLE/BURNUP with deterministic method for flux calculation and MVP-BURN with Monte Carlo method. The analysis results and discrepancies between different analysis methods are summarized in this paper. Sensibility studies of eigenvalue and sodium void reactivity of the medium size metallic fuel benchmark core are also conducted to determine the main reactions contributing to the difference between JENDL-4.0 and other libraries JEFF-3.1 and ENDF/B-VII.
    Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), Sep. 2014, [Reviewed]
  • Effects of nuclear data library and ultra-fine group calculation for large size sodium-cooled fast reactor OECD benchmarks
    Kugo Teruhiko; Sugino Kazuteru; Uematsu Mari Mariannu; Numata Kazuyuki*, The present paper summarizes calculation results for an international benchmark proposed under the framework of the Working Party on scientific issues of Reactor Systems (WPRS) of the Nuclear Energy Agency of the OECD. It focuses on the large size oxide-fueled SFR. Library effect for core performance characteristics and reactivity feedback coefficients is analyzed using sensitivity analysis. The effect of ultra-fine energy group calculation in effective cross section generation is also analyzed. The discrepancy is about 0.4\% for a neutron multiplication factor by changing JENDL-4.0 with JEFF-3.1. That is about -0.1\% by changing JENDL-4.0 with ENDF/B-VII.1. The main contributions to the discrepancy between JENDL-4.0 and ENDF/B-VII.1 are $^{240}$Pu capture, $^{238}$U inelastic scattering and $^{239}$Pu fission. Those to the discrepancy between JENDL-4.0 and JEFF-3.1 are $^{23}$Na inelastic scattering, $^{56}$Fe inelastic scattering, $^{238}$Pu fission, $^{240}$Pu capture, $^{240}$Pu fission, $^{238}$U inelastic scattering, $^{239}$Pu fission and $^{239}$Pu nu-value. As for the sodium void reactivity, JEFF-3.1 and ENDF/B-VII.1 underestimate by about 8\% compared with JENDL-4.0. The main contributions to the discrepancy between JENDL-4.0 and ENDF/B-VII.1 are $^{23}$Na elastic scattering, $^{23}$Na inelastic scattering and $^{239}$Pu fission. That to the discrepancy between JENDL-4.0 and JEFF-3.1 is $^{23}$Na inelastic scattering. The ultra-fine energy group calculation increases the sodium void reactivity by 2\%.
    Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), Sep. 2014, [Reviewed]
  • Development of a calculation system for the estimation of decontamination effect
    Satoh Daiki; Kojima Kensuke; Oizumi Akito; Matsuda Norihiro; Iwamoto Hiroki; Kugo Teruhiko; Sakamoto Yukio*; Endo Akira; Okajima Shigeaki, A calculation system for the estimation of decontamination effect (CDE) has been developed to support planning a rational and effective decontamination. The method calculates the dose-rate distribution before and after decontamination, according to the distribution of radioactivity and the decontamination factor (DF), and uses a dose rate reduction factor (DRRF) to estimate the decontamination effect. The results that were calculated by using the CDE were compared with the results of measurements as well as with the results of calculations that were performed using a Monte Carlo particle transport code PHITS. It was found that the CDE successfully reproduced the measured as well as the calculated dose-rate distributions, requiring less than several seconds of calculation time., Informa UK Limited
    Journal of Nuclear Science and Technology, May 2014, [Reviewed]
  • Benchmark calculations for reflector effect in fast cores by using the latest evaluated nuclear data libraries
    Fukushima Masahiro; Ishikawa Makoto; Numata Kazuyuki*; Jin Tomoyuki*; Kugo Teruhiko, Benchmark calculations for reflector effects in fast cores were performed to validate the reliability of scattering data of structural materials in the major evaluated nuclear data libraries, JENDL-4.0, ENDF/B-VII.1 and JEFF-3.1.2. The criticalities of two FCA and two ZPR cores were analyzed by using a continuous energy Monte Carlo calculation code. The ratios of calculation to experimental values were compared between these cores and the sensitivity analyses were performed. From the results, the replacement reactivity from blanket to SS and Na reflector is better evaluated by JENDL-4.0 than by ENDF/B-VII.1 mainly due to the $\bar{\mu}$ values of Na and $^{52}$Cr.
    Nuclear Data Sheets, Apr. 2014, [Reviewed]
  • Evaluation of neutron economical effect of new cladding materials in light water reactors
    Oizumi Akito; Akie Hiroshi; Iwamoto Nobuyuki; Kugo Teruhiko, Iron (Fe), nickel (Ni), titanium (Ti), niobium (Nb) and vanadium (V) are selected as possible component elements to cover a variety of new cladding materials for light water reactors (LWRs). The effect of larger thermal absorption cross sections of these elements than those of zirconium (Zr), together with those of silicon carbide (SiC), on the neutron economy in LWRs is evaluated by performing pin cell burnup calculations for a conventional pressurized water reactor (PWR), a low moderation high burnup LWR (LM-LWR) and a high moderation high burnup LWR (HM-LWR). As can be anticipated from the thermal cross sections, SiC has excellent neutron economy. The materials other than SiC largely decreases discharge burnup in comparison with Zircaloy (Zry). Among such elements of larger thermal absorption cross section, Nb has neutron economical advantage over the other materials except SiC in softer neutron spectrum reactors such as HM-LWR in which the atomic number ratio of hydrogen to heavy metal is 6. In the conventional LWRs, stainless steel of low Ni contents is as well as Nb for cladding material. The results of the analyses are summarized for the purpose to provide reference data for new cladding material development studies, in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zry cladding.
    Journal of Nuclear Science and Technology, Jan. 2014, [Reviewed]
  • Calculation system for the estimation of decontamination effect
    Satoh Daiki; Kojima Kensuke; Oizumi Akito; Matsuda Norihiro; Iwamoto Hiroki; Kugo Teruhiko; Sakamoto Yukio*; Endo Akira; Okajima Shigeaki, A computer software, named CDE (Calculation system for Decontamination Effect), has been developed to support planning the decontamination. CDE is programed with VBA (Visual Basic for Applications), and runs on Microsoft Excel with a user friendly graphical interface. It calculates dose rate distributions in a target area before and after the decontamination from a radioactivity distribution and DF (Decontamination Factor), which is a ratio of original radioactivity to remaining one after the decontamination. DRRF (Dose Rate Reduction Factor) is also derived to express the decontamination effect. All the calculation results are visualized on an image of the target area with color map. Owing to its quick calculation speed, CDE is able to investigate the decontamination effect in various cases for a short period. This is very useful to establish a rational decontamination plan before an action.
    Transactions of the American Nuclear Society, Nov. 2013, [Reviewed]
  • Major safety and operational concerns for fuel debris criticality control
    Tonoike Kotaro; Sono Hiroki; Umeda Miki; Yamane Yuichi; Kugo Teruhiko; Suyama Kenya, JAEA is conducting studies on criticality control of the fuel debris formed in the accident of Fukushima-Daiichi site. A new control principle must be established, referring principles for existing facilities, and based on criticality characteristics of the debris. In accordance with the principle, safe and practical control has to be realized for the debris whose condition is uncertain at present. This report outlines the present condition of debris and Fukushima site, introduces examples of criticality analysis, and discusses control principles. Research subjects are also proposed to realize the control.
    Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), Sep. 2013, [Reviewed]
  • Extended cross-section adjustment method to improve the prediction accuracy of core parameters
    Yokoyama Kenji; Ishikawa Makoto; Kugo Teruhiko, An extended cross-section adjustment method has been developed to improve the prediction accuracy of target core parameters. The present method is on the basis of a cross-section adjustment method which minimizes the uncertainties of target core parameters under the conditions that integral experimental data are given. The present method enables us to enhance the prediction accuracy better than the conventional cross-section adjustment method by taking into account the target core parameters, as well as the extended bias factor method. In addition, it is proved that the present method is equivalent to the extended bias factor method when only one target core parameter is taken into account. The present method is implemented in an existing cross-section adjustment solver. Numerical calculations verify the derived formulation and demonstrate an applicability of an adjusted cross-section set which is specialized for the target core parameters.
    Journal of Nuclear Science and Technology, Dec. 2012, [Reviewed]
  • Measurement and analysis of reflector reactivity worth by replacing stainless steel with zirconium at the fast critical assembly (FCA)
    Fukushima Masahiro; Kitamura Yasunori; Ando Masaki; Kugo Teruhiko, Zirconium alloy instead of stainless steel (SS) has been considered as an effective reflector to improve the neutron economy in the experimental fast reactor JOYO. The aim of the present study is to demonstrate the effectiveness of the zirconium (Zr) reflector compared with the SS reflector in a fast reactor core. The FCA-XXVIII-1(3) core was built at the fast critical facility (FCA) and the reflector reactivity worth was measured by replacing SS with Zr at the peripheral region of the core. The experimental result of the positive reflector reactivity worth demonstrates the effectiveness of the Zr reflector compared with the SS reflector in the fast reactor core. This paper also focuses on the validation of standard calculation methods used for fast reactors with JENDL-4.0. As a result, it is confirmed that the standard calculation methods for the reflector reactivity worth show agreement within the experimental error.
    Journal of Nuclear Science and Technology, Oct. 2012, [Reviewed]
  • Results of the program to develop guidelines for decontamination: (1) Planning of decontamination with use of a calculation system for decontamination effect (CDE)
    KUGO Teruhiko; MATSUDA Norihiro; OIZUMI Akito, 福島の環境修復に向けた福島県除染ガイドライン作成調査事業において、除染前後の空間線量率の変化を予測する除染効果評価システムCDEを活用した除染計画の策定結果について報告する。, Atomic Energy Society of Japan
    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan, 2012
  • Benchmark calculations of sodium-void experiments with uranium fuels at the fast critical assembly FCA
    Fukushima Masahiro; Kitamura Yasunori; Kugo Teruhiko; Yamane Tsuyoshi; Ando Masaki; Chiba Go; Ishikawa Makoto; Okajima Shigeaki, The capture cross section of $^{235}$U has been re-evaluated by the OECD/NEA/NSC/WPEC subgroup 29 focusing on energy region from 100 eV to 1 MeV from the viewpoints of differential and integral data analyses since 2007. Sodium-void reactivity experiments with uranium fuels were carried out at the Fast Critical Assembly (FCA) in the Japan Atomic Energy Agency (JAEA) in 2009 and new integral data were obtained to help to validate the re-evaluated capture cross section of $^{235}$U. The benchmark calculations for the new integral data were performed by using a continuous-energy Monte Carlo code (MVP) with use of the evaluated nuclear data libraries JENDL-3.2, -3.3, -4.0, ENDF/B-VII.0 and JEFF-3.1. The ratios of calculated to experimental (C/E) values of sodium-void reactivities with respect to JENDL-3.3, ENDF/B-VII.0 and JEFF-3.1 are less than those with respect to JENDL-3.2 and -4.0. The analysis results are similar to those of sodium-void reactivities previously obtained at the BFS facility in Russia. The benchmark calculations demonstrate the improvement of the reliability of the integral data such as the new integral data obtained at the FCA and the previously obtained data in the BFS and the usefulness of the new integral data for the validation of the re-evaluated cross section of $^{235}$U.
    Progress in Nuclear Science and Technology (Internet), Oct. 2011, [Reviewed]
  • Recent application of nuclear data to reactor core analysis in JAEA
    Kugo Teruhiko, The status of application study of nuclear data in JAEA is introduced in the following categories, theoretical and study and code development, experiment and extension of experimental database and application to reactor core analysis including the fuel cycle back-end field. Sensitivities of fission product concentrations of LWR burned fuel to nuclear data have been evaluated based on the depletion perturbation theory. Important nuclear data are specified for the accurate prediction of fission product concentrations. Extended bias factor methods are applied for the uncertainty evaluation of a fast breeder reactor core. From the application, the extended bias factor methods powerfully eliminate the uncertainty due to cross section errors by utilizing a number of integral data.
    Journal of the Korean Physical Society, Aug. 2011, [Reviewed]
  • Development of a unified cross-section set ADJ2010 based on adjustment technique for fast reactor core design
    Sugino Kazuteru; Ishikawa Makoto; Yokoyama Kenji; Nagaya Yasunobu; Chiba Go; Hazama Taira; Kugo Teruhiko; Numata Kazuyuki*; Iwai Takehiko*; Jin Tomoyuki*, 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.
    Journal of the Korean Physical Society, Aug. 2011, [Reviewed]
  • JENDL-4.0 integral testing for fission systems
    Okumura Keisuke; Sugino Kazuteru; Chiba Go; Nagaya Yasunobu; Yokoyama Kenji; Kugo Teruhiko; Ishikawa Makoto; Okajima Shigeaki, The new version of Japanese evaluated nuclear data library, JENDL-4.0, is tested with integral data of fission systems. This data testing is carried out with a wide range of integral data including the critical benchmarks preserved in the International Handbook of Evaluated Criticality Safety Benchmark Experiments, the experimental data of MOX-fueled critical assemblies relating to the plutonium aging effect, the critical data of various fast critical assemblies and the fast reactors JOYO and MONJU, and the post-irradiation examination data of the pressurized-water reactor Takahama-3 and the fast reactor JOYO. The benchmark calculations are performed with a continuous-energy Monte Carlo code MVP-II or a sophisticated deterministic neutron transport code system. Benchmark calculations with other libraries, such as JENDL-3.3, ENDF/B-VII.0 and JEFF-3.1, are also performed, and differences in performance of these libraries are discussed with a help of sensitivity profiles to nuclear data., Korean Physical Society
    Journal of the Korean Physical Society, Aug. 2011, [Reviewed]
  • Effect of polynomial expansion order of intranode flux treatment in nodal S$_{N}$ transport calculation code NSHEX for large-size fast power reactor core analysis
    Sugino Kazuteru; Kugo Teruhiko, The nodal discrete ordinates (S$_{N}$) transport calculation code for three-dimensional hexagonal geometry NSHEX treats intra-node flux distribution by polynomial series and considers angular dependence of flux by the S$_{N}$ method. For the improvement of calculation accuracy of NSHEX for the practical use to large-size fast reactor plants, the maximum order of polynomial series is extend from two to six. In order to check the effect of the polynomial expansion order, NSHEX is applied to the middle-size fast power reactor core "Monju" and the large-size one "Super Phenix" including various control rod insertion conditions. From the application, it is found that extension of polynomial expansion order is effective especially for the large-size core "Super Phenix" with control rod inserted condition., Atomic Energy Society of Japan
    Journal of Nuclear Science and Technology, Mar. 2011, [Reviewed]
  • JENDL-4.0 benchmarking for fission reactor applications
    Chiba Go; Okumura Keisuke; Sugino Kazuteru; Nagaya Yasunobu; Yokoyama Kenji; Kugo Teruhiko; Ishikawa Makoto; Okajima Shigeaki, 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 may become possible by using the library JENDL-4.0., Atomic Energy Society of Japan
    Journal of Nuclear Science and Technology, Feb. 2011, [Reviewed]
  • Design study of nuclear power systems for deep space explorers, 1; Criticality of low enriched uranium fueled core
    Kugo Teruhiko; Akie Hiroshi; Yamaji Akifumi; Nabeshima Kunihiko; Iwamura Takamichi; Akimoto Hajime, Combining a nuclear reactor with thermoelectric converters is expected to be one of promising options to supply a propulsion power for deep space explorers. One of the key features of the concept is to use low enriched uranium fuels from the viewpoint of nuclear non-proliferation. Fuels of uranium oxide, nitride and metal were examined. Zirconium and yttrium hydrides, beryllium, zirconium beryllide and graphite were considered as moderators. Reflectors of beryllium, beryllium oxide, zirconium beryllide and graphite were taken into consideration. A criticality survey of the core was performed by changing the ratio of the fuel, moderator and structure, and the reflector thickness. As a result from the viewpoint of a smaller mass of reactor, it is better to use thermal spectrum cores than fast ones, and the metal hydride moderators than beryllium or graphite. For example, a combination of uranium nitride, yttrium hydride and beryllium reflector achieves a reactor mass of as low as 500kg.
    Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), May 2009, [Reviewed]
  • Prediction accuracy improvement of neutronic characteristics of a breeding light water reactor core by extended bias factor methods with use of FCA-XXII-1 critical experiments
    Kugo Teruhiko; Ando Masaki; Kojima Kensuke; Fukushima Masahiro; Mori Takamasa; Nakano Yoshihiro; Okajima Shigeaki; Kitada Takanori*; Takeda Toshikazu*
    Journal of Nuclear Science and Technology, Apr. 2008, [Reviewed]
  • Application of bias factor method with use of exponentiated experimental value to prediction uncertainty reduction in coolant void reactivity of breeding light water reactor
    Kugo Teruhiko; Kojima Kensuke; Ando Masaki; Mori Takamasa; Takeda Toshikazu*, We have applied the bias factor method to coolant void reactivity of a breeding light water reactor with use of FCA-XXII-1 experiment with introducing a concept of exponentiated experimental value into the bias factor method in order to overcome a problem caused by the conventional bias factor method in which the prediction uncertainty increases in the case that the experimental core has the opposite reactivity worth and the consequent opposite sensitivity coefficients to the real core. In the present study, we have formulated the prediction uncertainty reduction by the use of the bias factor method extended by the concept of the exponentiated experimental value. From the numerical results, it is verified that the concept of exponentiated experimental value can improve the prediction accuracy compared with the original uncertainty in the design calculation value while the conventional bias factor method cannot improve the prediction accuracy. It is concluded that the introduction of exponentiated experimental value can effectively utilize experimental data and extend applicability of the bias factor method., The Japan Society of Mechanical Engineers
    Journal of Power and Energy Systems (Internet), Jan. 2008, [Reviewed]
  • Theoretical study on new bias factor methods to effectively use critical experiments for improvement of prediction accuracy of neutronic characteristics
    Kugo Teruhiko; Mori Takamasa; Takeda Toshikazu*, The extended methods with two concepts, the LC and the PE methods, are proposed to enhance the bias factor method for improvement of the prediction accuracy of neutronic characteristics. The two methods utilize a number of critical experimental results and produce a semi-fictitious experimental value with them. The LC method defines a bias factor by a ratio of a linear combination of experimental values to that of calculation values for the experiments. The PE method defines it by a ratio of a product of exponentiated experimental values to that of exponentiated calculation values. We formulate how to determine weights for the LC method and exponents for the PE method in order to minimize the variance of the design prediction value. From a theoretical comparison among the two methods, the conventional method and the previously proposed method called the generalized bias factor method, it is concluded that the PE method is the most useful method in order to improve the prediction accuracy. Main advantages of the PE method are summarized as follows. The prediction accuracy is necessarily improved compared with the design calculation value and is the most improved by utilizing all the experimental results. From these facts, it can be said that the PE method effectively utilizes all the experimental results and has a possibility to make a full scale mockup experiment unnecessary with use of existing and future benchmark experiments., Atomic Energy Society of Japan
    Journal of Nuclear Science and Technology, Dec. 2007, [Reviewed]
  • Application of bias factor method with use of virtual experimental value to prediction uncertainty reduction in void reactivity worth of breeding light water reactor
    Kugo Teruhiko; Mori Takamasa; Kojima Kensuke; Takeda Toshikazu*, Utilizing the critical experiments for MOX fueled tight lattice LWR cores at FCA XXII-1 cores, we have evaluated prediction uncertainty reduction in coolant void reactivity worth of a breeding LWR core based on the bias factor method. In the present study, to extend the applicability of the bias factor method, we have introduced an exponentiated experimental value as a virtual experimental value and formulated the prediction uncertainty reduction with the bias factor method extended by the concept. From the numerical evaluation, it has been shown that the prediction uncertainty due to cross section errors has been reduced by the use of the concept of the virtual experimental value. It is concluded that the introduction of virtual experimental value can effectively utilize experimental data and extend applicability of the bias factor method., The Japan Society of Mechanical Engineers
    Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), Apr. 2007, [Reviewed]
  • Conceptual design of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) and its recycle characteristics
    Uchikawa Sadao; Okubo Tsutomu; Kugo Teruhiko; Akie Hiroshi; Takeda Renzo*; Nakano Yoshihiro; Onuki Akira; Iwamura Takamichi, 軽水炉技術に立脚し、現行軽水炉燃料サイクルに適合したプルトニウム有効利用を実現し、将来的には同一炉心構成の下で増殖型への発展が可能な革新的水冷却炉概念(FLWR)を、低減速軽水炉概念を発展させて構築した。本論文では、軽水炉技術によるプルトニウム利用高度化の考え方,FLWRの基本構成と主要特性を報告する。, Atomic Energy Society of Japan
    Journal of Nuclear Science and Technology, Mar. 2007, [Reviewed]
  • Concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR)
    Iwamura Takamichi; Uchikawa Sadao; Okubo Tsutomu; Kugo Teruhiko; Akie Hiroshi; Nakano Yoshihiro; Nakatsuka Toru, In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming Mixed Oxide (MOX)-LWR technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Agency (JAEA). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming MOX-LWR technologies without significant technical gaps. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-developed LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the future fuel cycle circumstances during the reactor operation period around 60 years. Investigation on the core for both the parts of the FLWR concepts has been performed, including the core conceptual design, the core characteristics under Pu multiple recycling, the thermal hydraulic investigation in the tight-lattice core, and so forth. Up to the present, promising results have been obtained.
    Nuclear Engineering and Design, Aug. 2006, [Reviewed]
  • Investigation on Innovative Water Reactor for Flexible Fuel Cycle (FLWR), 1; Conceptual design
    Uchikawa Sadao; Okubo Tsutomu; Kugo Teruhiko; Akie Hiroshi; Nakano Yoshihiro; Onuki Akira; Iwamura Takamichi, 軽水炉技術に立脚し、現行軽水炉燃料サイクルに適合したプルトニウム有効利用を実現し、将来的には同一炉心構成の下で増殖型への発展が可能な革新的水冷却炉概念(FLWR)を、低減速軽水炉概念を発展させて構築した。本論文では、軽水炉技術によるプルトニウム利用高度化の考え方,FLWRの基本構成と主要特性、並び関連する要素技術の研究開発状況を報告する。
    Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), Oct. 2005, [Reviewed]
  • Investigation on Innovative Water Reactor for Flexible Fuel Cycle (FLWR), 2; Recycle characteristics
    Okubo Tsutomu; Uchikawa Sadao; Kugo Teruhiko; Akie Hiroshi; Takeda Renzo*, In order to ensure sustainable energy supply in the future based on the commercialized LWR technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in JAERI. Results on the FLWR recycling characteristics under possible various reprocessing schemes are presented in the present paper. The results show the recycling is possible a few times at most as long as the fissile Pu content stays over 60\%, even in the high conversion type core with the conversion ratio around 0.9, under the simplified PUREX reprocessing, with relatively high average decontamination factor. For breeding core, the results have indicated that even under the reprocessing with relatively low DFs and with whole MA, the recycling is also feasible, suggesting all MAs from the core can be possibly recycled itself, although the core performances are a little degraded depending on MA and FP contents.
    Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), Oct. 2005, [Reviewed]
  • Benchmark solution for unstructured geometry PWR problem by method of characteristics using combinatorial geometry
    Kugo Teruhiko; Mori Takamasa, A new deterministic transport code based on the method of characteristics (MOC) has been developed for heterogeneous transport calculations in core design of innovative reactors which have complex structures. We have investigated the capability of the MOC code for general geometry with an unstructured geometry PWR problem. The comparison of the results with accurate Monte Carlo calculation results by GMVP has confirmed that the MOC code produces satisfactory results and has a capability to treat unstructured geometry.
    Proceedings of International Topical Meeting on Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications (M\&C 2005) (CD-ROM), Sep. 2005, [Reviewed]
  • Concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR)
    Iwamura Takamichi; Uchikawa Sadao; Okubo Tsutomu; Kugo Teruhiko; Akie Hiroshi; Nakatsuka Toru, In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming LWR-Mixed Oxide (MOX) technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR without significant gaps in technical point of view. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the fuel cycle circumstances.
    Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), May 2005, [Reviewed]
  • Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor
    Shelley A.; Shimada Shoichiro*; Kugo Teruhiko; Okubo Tsutomu; Iwamura Takamichi, Parametric studies have been done for a PWR-type reduced-moderation water reactor (RMWR) with seed-blanket fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup. It was found that 50 to 60\% of seed in a seed-blanket assembly has higher conversion ratio. The number of seed-blanket layers is 20, in which the number of seed layers is 15 and blanket layers is 5. The fuel assembly with the height of seed of 1000mm$\times$2, internal blanket of 150 mm and axial blanket of 400mm$\times$2 is recommended. The conversion ratio is 1.0 and the average burnup in core region is 38.2 GWd/t. The enrichment of fissile Pu is 14.6 wt\%. The void coefficient is +21.8 pcm/\% void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account. It is also possible to use this fuel assembly for a high core averaged burnup of 45GWd/t, however, the height of seed must be 500mm$\times$2 to improve the void coefficient. The conversion ratio is 0.97 and void coefficient is +20.8 pcm/\%void.
    Nuclear Engineering and Design, Oct. 2003, [Reviewed]
  • Fast vector computation of the characteristics method
    Kugo Teruhiko, Two vector computation algorithms; an odd-even sweep (OES) method and an independent sequential sweep (ISS) method, have been developed for the characteristics method to solve the neutron transport equation in a heterogeneous geometry. They realize long vector lengths without recursive operations for effective use of vector computers. Their efficiency has been investigated to a realistic fuel assembly calculation. For both methods, a vector computation is 15 times faster than a scalar computation. From a viewpoint of a comparison between the OES and ISS methods, the ISS method is superior to the OES method because the ISS method shows a faster convergence and saves a computer memory without reducing a computation speed.
    Journal of Nuclear Science and Technology, Mar. 2002, [Reviewed]

MISC

Lectures, oral presentations, etc.