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

https://doi.org/10.1080/00223131.2017.1299644

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, Np, Pu, Pu, Pu, Am, Am, and 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 Pu. For all the libraries, the benchmark tests by a Monte Carlo calculation code show obvious overestimations particularly for the fission rate ratios of Cm to Pu. Additionally, a large discrepancy about by 20% between the libraries is revealed for the fission rate ratio of Pu to Pu measured in the intermediate neutron spectrum. The cause of discrepancy is furthermore clarified by sensitivity analyses.

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

https://doi.org/10.1051/epjconf/201611107002

In the IX-th series at FCA, central fission rate ratios for TRU such as Np, Pu, Pu, Am, Am, Cm and 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.

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 Np, Pu, Pu, Am, Am and Cm to 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.

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

https://doi.org/10.1080/00223131.2012.723173

Benchmark calculations of sodium-void experiments with uranium fuels at the fast critical assembly FCA for U capture cross section in the keV region

Fukushima, Masahiro

Kaku Deta Nyusu (Internet), (101), p.49 - 54, 2012/02

http://wwwndc.jaea.go.jp/JNDC/ND-news/pdf101/No101-07.pdf

no abstracts in English

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

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 transport code. Conventional calculations with a standard 70 energy group structure and under the P transport approximation overestimated the experimental values up to 1.5%/. 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%/ with a 140 energy group structure and under the P approximation.

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 as converter material because Th makes fission reaction with fast neutrons. The place where Th can be uses is limited by regulations because Th is nuclear fuel material. The purpose of this research is to develop a new optical fiber detector to measure fast neutrons without 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.

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

https://doi.org/10.1299/jpes.2.73

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.

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.