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Journal Articles

Benchmark models for criticalities of FCA-IX assemblies with systematically changed neutron spectra

Fukushima, Masahiro; Kitamura, Yasunori; Kugo, Teruhiko; Okajima, Shigeaki

Journal of Nuclear Science and Technology, 53(3), p.406 - 424, 2016/03

 Times Cited Count:9 Percentile:69.23(Nuclear Science & Technology)

Journal Articles

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

Annals of Nuclear Energy, 81, p.117 - 124, 2015/07

 Times Cited Count:2 Percentile:19.3(Nuclear Science & Technology)

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.

Journal Articles

Options of principles of fuel debris criticality control in Fukushima Daiichi reactors

Tonoike, Kotaro; Sono, Hiroki; Umeda, Miki; Yamane, Yuichi; Kugo, Teruhiko; Suyama, Kenya

Nuclear Back-end and Transmutation Technology for Waste Disposal, p.251 - 259, 2015/00

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.

Journal Articles

Applications of integral benchmark data

Palmiotti, G.*; Briggs, J. B.*; Kugo, Teruhiko; Trumble, E.*; Kahler, A. C.*; Lancaster, D.*

Nuclear Science and Engineering, 178(3), p.295 - 310, 2014/11

 Times Cited Count:8 Percentile:51.25(Nuclear Science & Technology)

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.

Journal Articles

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*

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 12 Pages, 2014/09

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%.

Journal Articles

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*

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 15 Pages, 2014/09

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.

Journal Articles

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.*; et al.

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 16 Pages, 2014/09

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.

JAEA Reports

Report on decontamination pilot projects to establish guidelines for environmental remediation of residential areas contaminated with radioactive materials discharged from the Fukushima Daiichi Nuclear Power Station Accident

Kihara, Shinji; Amazawa, Hiroya; Sakai, Akihiro; Nakata, Hisakazu; Kugo, Teruhiko; Matsuda, Norihiro; Oizumi, Akito; Sasamoto, Hiroshi; Mitsui, Seiichiro; Miyahara, Kaname

JAEA-Research 2013-033, 320 Pages, 2014/07


JAEA performed decontamination experiments at two test sites that combined a range of buildings and different types of land use, located in Date and Minami Soma municipalities as field pilot projects in order to accumulate knowledge and data for full-scale decontamination activities performed by local governments. In the pilot projects, we established its plan using practical decontamination methods that can be easily implemented, according to decontamination targets (e.g., forests, agricultural land, residential house and roads) at each site. As a result of the decontamination, the average air dose rates were reduced to approximately one half of the values before decontamination.

Journal Articles

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

Journal of Nuclear Science and Technology, 51(5), p.656 - 670, 2014/05

 Times Cited Count:7 Percentile:51.25(Nuclear Science & Technology)

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.

Journal Articles

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

Nuclear Data Sheets, 118, p.405 - 409, 2014/04

 Times Cited Count:0 Percentile:0.02(Physics, Nuclear)

JAEA Reports

Study to improve recriticality evaluation methodology after severe accident (Joint research)

Kugo, Teruhiko; Ishikawa, Makoto; Nagaya, Yasunobu; Yokoyama, Kenji; Fukaya, Yuji; Maruyama, Hiromi*; Ishii, Yoshihiko*; Fujimura, Koji*; Kondo, Takao*; Minato, Hirokazu*; et al.

JAEA-Research 2013-046, 53 Pages, 2014/03


The present report summarizes the results of a 2-year cooperative study between JAEA and Hitachi-GE in order to contribute to the settlement of the Fukushima-Daiichi Nuclear Power Plants which suffered from the severe accident on March 2011. In the present study, the possible scenarios to reach the recriticality events in Fukushima-Daiichi were investigated first. Then, the analytical methodology to evaluate the time-dependent recriticality events has been developed by modelling the reactivity insertion rate and the possible feedback according to the recriticality scenarios identified in the first step. The methodology developed here has been equipped as a transient simulation tool, PORCAS, which is operated on a multi-purpose platform for reactor analysis, MARBLE. Finally, the radiation exposure rates by the postulated recriticality events in Fukushima-Daiichi were approximately evaluated to estimate the impact to the public environment.

JAEA Reports

Database for nuclear data sensitivity of burnup composition in light water reactors

Oizumi, Akito; Jin, Tomoyuki*; Yokoyama, Kenji; Ishikawa, Makoto; Kugo, Teruhiko

JAEA-Data/Code 2013-019, 278 Pages, 2014/02


In design work for nuclear fuel cycle plants, decommissioning facilities and light water reactors (LWRs), it has been feasible to quantitatively evaluate the uncertainty of fuel burnup characteristics with identifying error sources arising from the analytical modeling or the related physical property such as nuclear data. Owing to the recent improvement of sensitivity analysis method and enhancement of computer capability, this new evaluation technology would be a promising strategy against the current demand for quality assurance, verification & validation (V&V) and accountability. The present report summarizes nuclear-data sensitivity of atomic number densities after burnup for the LWR fuels of UO$$_{2}$$ and MOX in PWR and BWR. The analysis method is based on the generalized perturbation theory with JENDL-4.0 and a multi-purpose reactor analysis code MARBLE. The present study focuses on 35 fission products and 18 actinides. Sensitivities are calculated with respect to multigroup cross sections, half-lives and fission yields. Electronic files of the sensitivities are stored in a compact disk as a database. Important trends of the sensitivities are presented and their physical mechanisms are discussed. By incorporating the sensitivities with nuclear data covariance and post irradiation examination data, it would be possible to meet the demand for V&V and to break down the uncertainty due to nuclear data into dominant error sources. Thus, the sensitivities can be used to suggest the needs for nuclear data measurements and to extract those for reactor physics experiments in order to make the strategic deliberation of design rationalization.

Journal Articles

Evaluation of neutron economical effect of new cladding materials in light water reactors

Oizumi, Akito; Akie, Hiroshi; Iwamoto, Nobuyuki; Kugo, Teruhiko

Journal of Nuclear Science and Technology, 51(1), p.77 - 90, 2014/01

 Times Cited Count:4 Percentile:33.4(Nuclear Science & Technology)

Journal Articles

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

Transactions of the American Nuclear Society, 109(1), p.1261 - 1263, 2013/11

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.

Journal Articles

Uncertainty evaluation for $$^{244}$$Cm production in spent fuel of light water reactor by using burnup sensitivity analysis

Oizumi, Akito; Yokoyama, Kenji; Ishikawa, Makoto; Kugo, Teruhiko

JAEA-Conf 2013-002, p.59 - 64, 2013/10

Journal Articles

Major safety and operational concerns for fuel debris criticality control

Tonoike, Kotaro; Sono, Hiroki; Umeda, Miki; Yamane, Yuichi; Kugo, Teruhiko; Suyama, Kenya

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.729 - 735, 2013/09

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.

Journal Articles

Extended cross-section adjustment method to improve the prediction accuracy of core parameters

Yokoyama, Kenji; Ishikawa, Makoto; Kugo, Teruhiko

Journal of Nuclear Science and Technology, 49(12), p.1165 - 1174, 2012/12

 Times Cited Count:17 Percentile:78.83(Nuclear Science & Technology)

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 Articles

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

Journal of Nuclear Science and Technology, 49(10), p.961 - 965, 2012/10

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Decontamination planning based on computer simulation code CDE

Satoh, Daiki; Oizumi, Akito; Matsuda, Norihiro; Kojima, Kensuke; Kugo, Teruhiko; Sakamoto, Yukio*; Endo, Akira; Okajima, Shigeaki

RIST News, (53), p.12 - 23, 2012/09

Decontamination planning based on a computer simulation code CDE is discussed in this paper. Large amount of radionuclides had been discharged to environment in the accident of the Tokyo Electronic Power Corporation Fukushima Dai-ichi Nuclear Power Plant. CDE has been developed to support planning the decontamination. From the present study, it is validated that the computer simulation is very useful to predict the effect of the scenario before actions, and to plan the decontamination.

JAEA Reports

Development of calculation system for decontamination effect, CDE

Satoh, Daiki; Kojima, Kensuke; Oizumi, Akito; Matsuda, Norihiro; Kugo, Teruhiko; Sakamoto, Yukio*; Endo, Akira; Okajima, Shigeaki

JAEA-Research 2012-020, 97 Pages, 2012/08


A computer software, named CDE (Calculation system for Decontamination Effect), has been developed to support planning the decontamination. CDE calculates the dose rates before the decontamination by using a database of dose contributions by radioactive cesium. The decontamination factor is utilized in the prediction of the dose rates after the decontamination, and dose rate reduction factor is evaluated to express the decontamination effect. The results are visualized on the image of a target zone with color map. In this paper, the overview of the software and the dose calculation method are reported. The comparison with the calculation results by a three-dimensional radiation transport code PHITS is also presented. In addition, the source code of the dose calculation program and user's manual of CDE are attached as appendices.

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