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JAEA Reports

Annual report of Nuclear Science Research Institute, JFY2013 & 2014

Nuclear Science Research Institute

JAEA-Review 2018-036, 216 Pages, 2019/03

JAEA-Review-2018-036.pdf:19.22MB

Nuclear Science Research Institute (NSRI) is composed of Planning and Coordination Office, Fukushima Project Team and six departments, namely Department of Operational Safety Administration, Department of Radiation Protection, Engineering Services Department, Department of Research Reactor and Tandem Accelerator, Department of Fukushima Technology Development and Department of Decommissioning and Waste Management, and each departments manage facilities and develop related technologies to achieve the "Middle-term Plan" successfully and effectively. In order to contribute the future research and development and to promote management business, this annual report summarizes information on the activities of NSRI of JFY 2013 and 2014 as well as the activity on research and development carried out by Nuclear Safety Research Center, Advanced Research Center, Nuclear Science and Engineering Center and Quantum Beam Science Center, and activity of Nuclear Human Resource Development Center, using facilities of NSRI.

Journal Articles

Analyses with latest major nuclear data libraries of the fission rate ratios for several TRU nuclides in the FCA-IX experiments

Fukushima, Masahiro; Tsujimoto, Kazufumi; Okajima, Shigeaki

Journal of Nuclear Science and Technology, 54(7), p.795 - 805, 2017/07

 Times Cited Count:4 Percentile:36.15(Nuclear Science & Technology)

A series of integral experiments was conducted in FCA assemblies with systematically changed neutron spectra covering from the intermediate to fast ones. The experiments provide systematic data of central fission rates for TRU nuclides containing minor actinides, $$^{237}$$Np, $$^{238}$$Pu, $$^{239}$$Pu, $$^{242}$$Pu, $$^{241}$$Am, $$^{243}$$Am, and $$^{244}$$Cm. Latest major nuclear data libraries, JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2, were tested using benchmark models regarding the fission rate ratios relative to $$^{239}$$Pu. For all the libraries, the benchmark tests by a Monte Carlo calculation code show obvious overestimations particularly for the fission rate ratios of $$^{244}$$Cm to $$^{239}$$Pu. Additionally, a large discrepancy about by 20% between the libraries is revealed for the fission rate ratio of $$^{238}$$Pu to $$^{239}$$Pu measured in the intermediate neutron spectrum. The cause of discrepancy is furthermore clarified by sensitivity analyses.

Journal Articles

Benchmark tests of newly-evaluated data of $$^{235}$$U for CIELO project using integral experiments of uranium-fueled FCA assemblies

Fukushima, Masahiro; Kitamura, Yasunori*; Yokoyama, Kenji; Iwamoto, Osamu; Nagaya, Yasunobu; Leal, L. C.*

Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.605 - 619, 2016/05

A nuclear data of $$^{235}$$U has been recently evaluated for the CIELO (Collaborative International Evaluated Library Organization) project. We tested the newly-evaluated data of $$^{235}$$U using integral experiments of the Fast Critical Assembly (FCA) performed at JAEA. We selected two integral data of uranium-fueled FCA assemblies; one is the sodium-void reactivity worth of FCA XXVII-1 assembly and the other is the criticalities of the seven assemblies of FCA IX. The benchmark tests support the evaluation done in the resonance regions. However, the $$^{235}$$U capture cross section above the unresolved resonance range needs further investigation.

Journal Articles

Fission rate ratios of FCA-IX assemblies as integral experiment for assessment of TRU's fission cross sections

Fukushima, Masahiro; Tsujimoto, Kazufumi; Okajima, Shigeaki

EPJ Web of Conferences (Internet), 111, p.07002_1 - 07002_5, 2016/03

 Times Cited Count:2 Percentile:8.18

In the IX-th series at FCA, central fission rate ratios for TRU such as $$^{237}$$Np, $$^{238}$$Pu, $$^{242}$$Pu, $$^{241}$$Am, $$^{243}$$Am, $$^{244}$$Cm and $$^{239}$$Pu were measured in the seven uranium-fueled assemblies with systematically changed neutron spectra. The FCA-IX assemblies were constructed with simplicity both in geometry and composition. Taking their advantage, benchmark models with respect to central fission rate ratios had been recently established for the assessment of the TRU's fission cross sections. By virtue of these FCA-IX assemblies where the simple combinations of uranium fuel and diluent in their core regions were systematically varied, the neutron spectra of these benchmark models cover those of various reactor types, from fast to epithermal reactors. As an application of these benchmark models, JENDL-4.0 was utilized by a Monte Carlo calculation code. Several results show obvious discrepancies between calculation and experimental values. The benchmark models would be well suited for the assessment and improvement of the nuclear data for TRU's fission cross sections.

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:7 Percentile:22.3(Nuclear Science & Technology)

JAEA Reports

Establishment of benchmark problems for TRU fission rate ratios of FCA-IX assemblies

Fukushima, Masahiro; Oizumi, Akito; Iwamoto, Hiroki; Kitamura, Yasunori

JAEA-Data/Code 2014-030, 50 Pages, 2015/03

JAEA-Data-Code-2014-030.pdf:10.34MB

In the IX-th experimental series in 1980's at the fast critical assembly (FCA) facility, central fission rate ratios for TRU such as $$^{237}$$Np, $$^{238}$$Pu, $$^{242}$$Pu, $$^{241}$$Am, $$^{243}$$Am and $$^{244}$$Cm to $$^{239}$$Pu were measured in the seven uranium-fueled assemblies with systematically changed neutron spectra. In the present report, benchmark problems with respect to central fission rate ratios were established for the assessment of the TRU's fission cross sections. We reported the sample calculation results on the benchmark problems by using JENDL-4.0.

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:100(Physics, Nuclear)

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:100(Nuclear Science & Technology)

Journal Articles

Validation of fast reactor analysis methods using reactor physics experiments

Ando, Masaki

Nippon Genshiryoku Gakkai Dai-44-Kai Robutsuri Kaki Semina Tekisuto, p.139 - 163, 2012/08

Mock-up experiments for "JOYO" and "MONJU" using FCA and the JUPITER program using ZPPR are introduced. As a validation of the fast reactor analysis method, a process to make a benchmark model, examples of reactivity experiment (Doppler effect and sodium void effect) and benchmark experiments to evaluate nuclear data libraries are described.

Journal Articles

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

Progress in Nuclear Science and Technology (Internet), 2, p.306 - 311, 2011/10

Journal Articles

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

Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 6 Pages, 2010/10

Journal Articles

Development of the 4S and related technologies, 7; Summary of the FCA XXIII experiment analyses towards evaluation of prediction accuracies for the 4S core characteristics

Ueda, Nobuyuki*; Fukushima, Masahiro; Okajima, Shigeaki; Takeda, Toshikazu*; Kitada, Takanori*; Nauchi, Yasushi*; Kinoshita, Izumi*; Matsumura, Tetsuo*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9493_1 - 9493_9, 2009/05

A series of critical experiments were carried out in the JAEA fast critical facility (FCA) named FCA XXIII cores with placing emphases on the reflector reactivity worth and the sodium void reactivity which are especially important from the view point of safety features of the 4S. The analyses of those physics mockup experiments have been carried out by the neutron transport calculation methods with continuous energy Monte Carlo code MVP and 70 energy-group discrete ordinate P0-S8 transport code DANTSYS using libraries processed from JENDL-3.3 data file. The results showed that combination of the stochastic and deterministic transport calculation methods (Monte Carlo and Sn) provided good prediction bases for criticality, reflector worth, sodium void reactivity, reaction rate ratios and absorber reactivity worth for the 4S nuclear design.

Journal Articles

Experiment and analysis for criticality in small fast reactor with reflector at FCA

Fukushima, Masahiro; Okajima, Shigeaki; Mori, Takamasa; Takeda, Toshikazu*; Kinoshita, Izumi*

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 7 Pages, 2008/09

The criticality was measured in a series of mock-up cores simulated small fast reactor with massive reflector at FCA facility of JAEA in order to evaluate the prediction accuracy of the current analysis code system. In the analyses, the effective cross sections were obtained by using an ultra-fine group cell calculation code. The JENDL-3.3 cross section library was used. The core calculations for the criticality were performed by using a three-dimensional S$$_{N}$$ transport code. Conventional calculations with a standard 70 energy group structure and under the P$$_{0}$$ transport approximation overestimated the experimental values up to 1.5%$${Delta}$$$$k$$/$$k$$. Furthermore, the calculation parameters were investigated concerning the fine energy group structure and the higher Legendre order of anisotropic scattering cross section. Consequently, the calculation accuracy for the criticality was improved by about 1%$${Delta}$$$$k$$/$$k$$ with a 140 energy group structure and under the P$$_{3}$$ approximation.

Journal Articles

Benchmark test for TRU nuclear data by analysis of central fission rate ratios measured at FCA cores

Okajima, Shigeaki; Fukushima, Masahiro; Mukaiyama, Takehiko*

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 7 Pages, 2008/09

To validate the reliability of nuclear data of transuranium (TRU) in the evaluated nuclear data libraries, JENDL-3.3, ENDF/B-7.0 and JEFF-3.1, a benchmark test was performed by analyzing a series of central fission rate ratios of these nuclides (CFRR) measured at the Fast Critical Assembly (FCA) of JAEA. In the test a Monte Carlo calculation code is used. C/E values are compared between these data libraries.

Journal Articles

Development of optical fiber detector for measurement of fast neutron

Yagi, Takahiro*; Misawa, Tsuyoshi*; Pyeon, C. H.*; Unesaki, Hironobu*; Shiroya, Seiji*; Kawaguchi, Shinichi*; Okajima, Shigeaki; Tani, Kazuhiro*

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8 Pages, 2008/09

In order to insert a neutron detector in a narrow space such as a gap of between fuel plates and measure the fast neutrons in real time, a neutron detector with an optical fiber has been developed. This detector consists of an optical fiber whose tip is covered with mixture of neutron converter material and scintillator such as ZnS(Ag). The detector for fast neutrons uses ThO$$_{2}$$ as converter material because $$^{232}$$Th makes fission reaction with fast neutrons. The place where $$^{232}$$Th can be uses is limited by regulations because $$^{232}$$Th is nuclear fuel material. The purpose of this research is to develop a new optical fiber detector to measure fast neutrons without $$^{232}$$Th and to investigate the characteristic of the detector. These detectors were used to measure a D-T neutron generator and fast neutron flux distribution at Fast Critical Assembly. The results showed that the fast neutron flux distribution of the new optical fiber detector with ZnS(Ag) was the same as it of the activation method, and the detector are effective for measurement of fast neutrons.

Journal Articles

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, 45(4), p.288 - 303, 2008/04

 Times Cited Count:6 Percentile:52.62(Nuclear Science & Technology)

The effectiveness of the extended bias factor methods, the LC and PE methods, is numerically investigated by applying them to a breeding light water reactor core as a target core with use of FCA-XXII-1 critical experiments. The present study numerically verifies the features of the extended bias factor methods. Both the methods can improve the prediction accuracy the most by using all the experiments. The PE method always improves the prediction accuracy with any combination of experiments. The PE method is always superior to the LC method for improvement of the prediction accuracy. From the present study, the followings are found. The experiments on multiplication factor are more applicable to a reaction rate ratio of $$^{238}$$U capture to $$^{239}$$Pu fission (C28/F49) of the target core than the experiments on C28/F49. Combinations of the experiments on multiplication factor is more effective to a void reactivity of the target core than those of the experiments on void reactivity though those on void reactivity are superior to those on multiplication factors in the case of using a single experiment. From these results, we conclude that the experiments on multiplication factor are more effective than the other experiments for all the neutronic characteristics of the target core. From these results, it is concluded that the PE method is promising to complement full mockup experiments for various future nuclear systems by using a number of existing and future benchmark experiments.

Journal Articles

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*

Journal of Power and Energy Systems (Internet), 2(1), p.73 - 82, 2008/00

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.

Journal Articles

Measurement and analysis of $$^{238}$$U doppler reactivity effect in FCA cores simulating light-water-moderated MOX fuel lattices

Ando, Masaki; Fukushima, Masahiro; Okajima, Shigeaki; Kawasaki, Kenji*

Journal of Nuclear Science and Technology, 44(4), p.537 - 547, 2007/04

 Times Cited Count:1 Percentile:86.98(Nuclear Science & Technology)

$$^{238}$$U Doppler reactivity effect was measured using the FCA facility for the purpose of obtaining the data of the $$^{238}$$U Doppler reactivity effect in light-water-moderated MOX fuel and evaluating the prediction accuracy of current analysis code systems and nuclear data library. The experimental data of the Doppler reactivity effect from room temperature up to 800$$^{circ}$$C were obtained using various cylindrical natural-uranium samples in the mockup cores for MOX-fueled LWR with different neutron energy spectra and in a uranium fueled core. The analyses were performed using current standard analysis code systems for fast and thermal reactors with JENDL-3.3 data library. Both the analyses yielded calculated/experimental (C/E) ratios of 0.96 to 1.06 for the MOX cores, which showed a good agreement with the experiment within the experimental error and were similar to those in the uranium core.

Journal Articles

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*

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04

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.

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