Usami, Shin; Kishimoto, Yasufumi*; Taninaka, Hiroshi; Maeda, Shigetaka
JAEA-Technology 2018-003, 97 Pages, 2018/07
The decay heat used for effectiveness evaluation of the prevention measures against severe accidents in the prototype fast breeder reactor Monju was evaluated by applying the updated nuclear data libraries based on JENDL-4.0, reflecting the realistic core operation pattern, and setting the rational extent of uncertainty. The decay heats of fission products, the actinide nuclides such as Cm-242, and radioactive structural materials were calculated by FPGS code. The decay heat of U-239 and Np-239 was evaluated based on ANSI/ANS-5.1-1994. The calculation uncertainty of each decay heat was evaluated based on summation of uncertainty factors, C/E values of reaction rates obtained in Monju system startup test, and so on. Furthermore, the decay heat evaluation method based on the FPGS90 was verified by the comparison of the results of the decay heat measurement of the two spent MOX fuel subassemblies in the experimental fast reactor Joyo MK-II core.
Yamamoto, Takahiro; Ito, Chikara; Maeda, Shigetaka; Ito, Hideaki; Sekine, Takashi
JAEA-Technology 2017-036, 41 Pages, 2018/02
In the experimental fast reactor Joyo, the damaged upper core structure (UCS) was retrieved into the cask in May 2014 The dose rate on UCS surface was quite high due to the activation for over 30 years operation. In order to attain the optimum safety design, manufacture and operation of equipment for UCS replacement, the method to evaluate UCS surface dose rate was developed on the basis of C/E obtained by the in-vessel dose rate measurement in Joyo. In order to verify the evaluation method, the axial gamma-ray distribution measurement on the surface of the cask, which contained UCS, was conducted using a plastic scintillating optical fiber (PSF) detector. This paper describes the comparison results between calculation and measurement as follows. (1) The measured axial gamma-ray distribution on the cask surface had a peak on proper location with considering the cask shielding structure and agree well with the calculated distribution. (2) The C/E of axial gamma-ray distribution on the cask surface was ranged from 1.1 to 1.7. It was confirmed that the calculation for UCS replacement equipment design had a margin conservatively. Then, the results showed that the developed evaluation method for UCS replacement equipment design was sufficiently reliable.
Shiba, Tomooki; Maeda, Shigetaka; Sagara, Hiroshi*; Ishimi, Akihiro; Tomikawa, Hirofumi
Energy Procedia, 131, p.250 - 257, 2017/12
In the present paper, the ray source data was developed for the debris composition based on "best estimates", and the subsequent photon transportation calculation was performed to evaluate the leakage ray spectra according to the fuel debris. Since the creation of the line spectrum source requires a great deal, we have developed the relatively simple but accurate enough method to build up ray source, coupling of baseline spectra evaluated by ORIGEN2 code and several line spectra of interest. One of the advantages of the method is taking bremsstrahlung X rays into consideration by utilizing the bremsstrahlung libraries of ORIGEN2. The new ray source was used to calculate the detector response of HPGe detector and the results was compared as a benchmark with experimental measurement results of irradiated fuel pins. As the result, the simulated ray spectrum shape agreed well with the shape of ray spectrum obtained by the experiment.
Usami, Shin; Kishimoto, Yasufumi; Taninaka, Hiroshi; Maeda, Shigetaka
Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.3263 - 3274, 2016/05
The present paper describes the validation of the new decay heat evaluation method using FPGS90 code with both the updated nuclear data library and the rational extent of uncertainty, by comparing the results of the decay heat measurement of the spent fuel subassemblies in Joyo MK-II core and by comparing with the calculation results of ORIGEN2.2 code. The calculated values of decay heat (C) by FPGS90 based on the JENDL-4.0 library were coincident with the measured ones (E) within the calculation uncertainties, and the C/E ranged from 1.01 to 0.93. FPGS90 evaluated the decay heat almost 3% larger than ORIGEN2.2, and it improved the C/E in comparison with the ORIGEN2.2 code. Furthermore, The C/E by FPGS90 based on the JENDL-4.0 library was improved than that based on the JENDL-3.2 library, and the contribution of the revision of reaction cross section library to the improvement was dominant rather than that of the decay data and fission yield data libraries.
Takahashi, Naoki; Yoshinaka, Kazuyuki; Harada, Akio; Yamanaka, Atsushi; Ueno, Takashi; Kurihara, Ryoichi; Suzuki, Soju; Takamatsu, Misao; Maeda, Shigetaka; Iseki, Atsushi; et al.
Nippon Genshiryoku Gakkai Homu Peji (Internet), 64 Pages, 2016/00
no abstracts in English
Ota, Katsu; Ushiki, Hiroshi*; Maeda, Shigetaka; Kawahara, Hirotaka; Takamatsu, Misao; Kobayashi, Tetsuhiko; Kikuchi, Yuki; Tobita, Shigeharu; Nagai, Akinori
JAEA-Technology 2015-026, 180 Pages, 2015/11
In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of "MARICO-2" (material testing rig with temperature control) had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS). The replacement of the UCS was conducted from May to December 2014. The design and manufacture of UCS was started from 2008, and the installation of UCS was completed successfully in November 21th 2014. The major results gained during the design and manufacture of UCS is as follows.
Konno, Chikara; Maeda, Shigetaka; Kosako, Kazuaki*
Energy Procedia, 71, p.213 - 218, 2015/05
We report a problem on multigroup cross section data MATXS files with multiple temperatures. This problem was newly found out through neutron and flux calculations in a simple model of experimental fast reactor Joyo with DORT and MATXSLIB-J40, which is a multigroup cross section data file (300, 600, 900, 1200, 1800 K) of the latest Japanese Nuclear Data Library version 4.0 (JENDL-4.0) processed with the NJOY99 code. The calculated total neutron fluxes were almost the same both in 300 K and 600 K, while the total fluxes in 600 K were by 10% higher those that in 300 K. Through our detailed investigation, it was found out that the MATXS data format processed with NJOY was not consistent to that assumed in TRANSX for production data. In order to solve this problem, we made a simple program for modifying MATXS files to ones suitable to TRANSX. MATXSLIB-J40 will be revised with this program.
Ishihara, Kohei*; Takagi, Keisuke*; Minato, Haruna*; Kawarabayashi, Jun*; Tomita, Hideki*; Maeda, Shigetaka; Naka, Tatsuhiro*; Morishima, Kunihiro*; Nakano, Toshiyuki*; Nakamura, Mitsuhiro*; et al.
Radiation Measurements, 55, p.79 - 82, 2013/08
In order to measure the neutron under a condition of high intensity of -ray background, we made new nuclear emulsion based on non-sensitized OPERA emulsion which had small AgBr grain size (AgBr grain size of 60, 90 and 160 nm). The sensitivity of this new emulsion, which was a correlation between stopping power and grain density, was estimated experimentally by irradiating neutrons with several energies. We also simulated the response to -ray induced electrons and compared with some experimental results by using Co source. The results showed that there might be a threshold energy deposited in one AgBr grain under which it was impossible to develop. We estimated efficiency to the -ray and the neutron with this obtained response of the new emulsion.
Maeda, Shigetaka; Iguchi, Tetsuo*
Journal of Nuclear Science and Technology, 50(4), p.381 - 386, 2013/03
We present a new spectrum unfolding code, the Maximum Entropy and Maximum Likelihood Unfolding Code (MEALU), based on the maximum likelihood method combined with the maximum entropy method, which can determine a neutron spectrum without requiring an initial guess spectrum. The Normal or Poisson distributions can be used for the statistical distribution. MEALU can treat full covariance data for a measured detector response and response function. The algorithm was verified through an analysis of mock-up data and its performance was checked by applying it to measured data. The results for measured data from the experimental fast reactor Joyo also were compared with those obtained by the conventional J-log method for neutron spectrum adjustment. It was found that MEALU has potential advantages over conventional methods with regard to preparation of a priori information and error estimation.
Shiba, Tomooki*; Sagara, Hiroshi*; Onishi, Takashi; Koyama, Shinichi; Maeda, Shigetaka; Han, C. Y.*; Saito, Masaki*
Annals of Nuclear Energy, 51, p.74 - 80, 2013/01
The design consideration of DU-Am oxide fuel pin was performed for Pu denaturing in the framework of the protected plutonium production based on the irradiation analyses of the depleted U (DU) samples irradiated in the environment of broad range of neutron energy in the experimental fast reactor Joyo. From the results of irradiation analyses of DU, it was confirmed that there is a strong dependence of transmutation behavior of DU on the resonance neutron ratio even in a fast reactor. Also, it was confirmed that there is a strong effect of sample material form and shape on generated Pu nuclide inventory especially near the reflector area (20% resonance neutron ratio), because of the intensive self-shielding of U, though less is expected for Am. Sensitivity study of hypothetical DU-Am oxide fuel pellet irradiation on neutron energy and burn-up was performed to evaluate significant gradient of radial Pu isotopic composition profile (e.g., from 12 to 18% distribution in 3% Am doping, in 30% resonance neutron ratio and in 4.010 [n/cm] of neutron fluence inside a large pellet with softened neutron spectrum), and vulnerability of the fuel pellet surface in terms of Pu denaturing was revealed. Design consideration of radial zoning of Am content was introduced to flatten the radial distribution of isotopic composition of Pu. The results of radial zoning of Am (4% and 3% of Am in the outer and inner zone of DU-Am oxide fuel pellet) in hypothetical irradiation neutronics analysis showed the radial profile of produced Pu is over 15 at.% of Pu isotopic composition in any zone and meets the criteria of Kimura et al. based on decay heat of Pu to impede utilization to fission explosive devices.
Maeda, Shigetaka; Ito, Chikara; Sekine, Takashi; Aoyama, Takafumi
Journal of Power and Energy Systems (Internet), 6(2), p.184 - 196, 2012/06
The verification of calculated core characteristics of the Joyo MK-III core using the JUPITER fast reactor standard analysis method was conducted by comparing with the measured values through the core physics tests. The purpose is to upgrade the core performance to increase the driver fuel burn-up and to increase the excess reactivity necessary for conducting various irradiation tests in the core region. Most of the C/Es are within 5% of unity. Through the comparisons, the calculation accuracy of the JUPITER standard analysis method for a small size sodium cooled fast reactor with a hard neutron spectrum was clarified. As a result of this study, more irradiation tests can be performed with appropriate safety margin and the efficient core and fuel management can be achieved to save the number of refueling.
Kawarabayashi, Jun*; Ishihara, Kohei*; Takagi, Keisuke*; Tomita, Hideki*; Iguchi, Tetsuo*; Naka, Tatsuhiro*; Morishima, Kunihiro*; Maeda, Shigetaka
Journal of ASTM International (Internet), 9(3), 5 Pages, 2012/03
In order to measure the neutron from a spent fuel assembly in fast breeder reactor precisely, we made new nuclear emulsion based on non-sensitized OPERA film with AgBr grain size of 60, 90 and 160 nm. The efficiency for Cf neutron of the new emulsion was calculated to be 0.710 which energy ranged from 0.3 to 2 MeV that agrees with preliminary estimated value from experimental results. The sensitivity of the new emulsion was also estimated experimentally by irradiating 565 KeV and 14 MeV neutrons and found that the emulsion with the AgBr grain size of 60 nm showed the lowest sensitivity among these three emulsions but still had enough sensitivity for proton. Also, there was a suggestion from the experimental data that there was a threshold LET of 15 KeV/m for our new emulsion below which no silver cluster was developed. Further development of the response of nuclear emulsion with a few tens of nano-meter AgBr size is next step of this study.
Furukawa, Tomohiro; Kato, Shoichi; Maeda, Shigetaka; Yamamoto, Masaya; Sekine, Takashi; Ito, Chikara
JAEA-Research 2011-039, 20 Pages, 2012/02
Application of zirconium alloy as a neutron reflector around the driver fuel region of the Japanese experimental fast reactor JOYO has been planned for a further increase of core average burn-up. In order to investigate the compatibility of the zirconium alloys with high-temperature sodium which is coolant of the JOYO, corrosion test in sodium and tensile test of the exposed alloys were performed. The corrosion test was done at 500C and 650C in stagnant/flowing sodium for two kinds of zirconium alloys, and then weight change measurement and metallurgical observation were carried out. The tensile test was performed in air at the same temperature with the sodium exposure.
Maeda, Shigetaka; Ito, Chikara; Sekine, Takashi; Aoyama, Takafumi
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
The verification of calculated core characteristics of the Joyo MK-III core using the JUPITER fast reactor standard analysis method was conducted by comparing with the measured values through the core physics tests. The purpose is to upgrade the core performance to increase the driver fuel burn-up and to increase the excess reactivity necessary for conducting various irradiation tests in the core region. Most of the C/Es are within 5% of unity. Through the comparisons, the calculation accuracy of the JUPITER standard analysis method for a small size sodium cooled fast reactor with a hard neutron spectrum like Joyo was clarified. As a result of this study, more irradiation tests can be performed with appropriate safety margin and the efficient core and fuel management can be achieved to save the number of refueling.
Maeda, Shigetaka; Iguchi, Tetsuo*
Nippon Genshiryoku Gakkai Wabun Rombunshi, 10(2), p.63 - 75, 2011/06
Neutron spectrum unfolding is a widely applied technique to characterize neutron fields for various types of reactor dosimetry, where the neutron spectrum is derived from integral measured data such as multiple-foil activation rates, moderated neutron detector counts, etc. A lot of spectrum unfolding codes have been developed so far and their performances compared to each other. However, standardized metrology for neutron spectrum unfolding is not satisfactorily established yet from the viewpoint of adequate selection and usage of unfolding codes, response function database and input data set preparation. This article reviews the present status on the neutron spectrum unfolding techniques mainly related to the reactor dosimetry with activation foils and discusses the validity of the solution spectra obtained from different kinds of unfolding codes under a typical fast reactor neutron field. The results show that the solution spectrum strongly depends on a guess spectrum required to the input data as well as the theoretical assumption in each unfolding code. The issues to improve the accuracy of reactor dosimetry are summarized on the input spectrum, nuclear database and the standardization of unfolding procedure, respectively.
Maeda, Shigetaka; Yamamoto, Masaya; Soga, Tomonori; Sekine, Takashi; Aoyama, Takafumi
Journal of Nuclear Science and Technology, 48(4), p.693 - 700, 2011/04
Core modification was investigated to further increase the core burn-up of the experimental fast reactor Joyo. This modification also enables the core to accommodate more irradiation test subassemblies that have lower fissile material content compared to the driver fuel. The design calculations showed that the replacement of the radial reflector elements made of stainless steel with those made of zirconium of nickel-base ally is effective in improving neutron efficiency. The irradiation tests capacity can be increased by reducing the number of control rods based on the re-evaluation of the design margin in the control rod worth calculation. These modifications will be useful to save driver fuels and to enhance the Joyo's irradiation capability.
Ito, Hideaki; Maeda, Shigetaka; Naito, Hiroyuki; Akiyama, Yoichi; Miyamoto, Kazuyuki; Ashida, Takashi; Noguchi, Koichi; Ito, Chikara; Aoyama, Takafumi
JAEA-Technology 2010-049, 129 Pages, 2011/03
The in-vessel gamma dose rate was measured in the experimental fast reactor Joyo to evaluate the activation of reactor structural components and the radiation exposure of the fiber scope used for in-vessel visual inspection. The measurement system, which requires a wide sensitivity range and high durability in a high-temperature environment, was specifically developed for use in the sodium cooled fast reactor. Using this system, the in-vessel gamma dose rate with cooling times of 450 and 720 days after reactor shutdown was measured in Joyo, which has been operated for 71,000 hours over approximately 30 years. The gamma dose rate was calculated using QAD-CGGP2 code with the gamma source intensity obtained by the ORIGEN2 code. The neutron flux used as input to the ORIGEN2 was evaluated by the Joyo dosimetry method. The ratio between the calculated and experimental values ranged from 1.1 to 2.4, confirming the accuracy of gamma dose rate and component activation calculation.
Maeda, Shigetaka; Naito, Hiroyuki; Ito, Chikara; Aoyama, Takafumi
Progress in Nuclear Science and Technology (Internet), 1, p.182 - 185, 2011/02
In-vessel rate measurements have been conducted in the experimental fast reactor Joyo to obtain experimental data and to verify the analysis method. The in-vessel dose with cooling times of 500 and 700 days after reactor shutdown was measured in Joyo which had been operated for approximately 30 years. The dose was calculated using QAD code with the source obtained by ORIGEN2 code. The ratios between the calculated and experimental value was ranged 1.3 and 2.7. The accuracy of analys method for the amount and distribution of radioactive products and ray dose inside reactor vessel was confirmed for the in-vessel inspection or repair.
Maeda, Shigetaka; Ito, Chikara; Aoyama, Takafumi; Maeda, Yukimoto; Chatani, Keiji
Transactions of the American Nuclear Society, 103(1), p.581 - 582, 2010/11
The experimental fast reactor Joyo of the Japan Atomic Energy Agency is the first liquid sodium fast reactor in Japan. Thirty years of successful operation of Joyo has shown excellent safety and reliability, and has contributed much to the LMFBR development program. Many kinds of irradiation experience have been accumulated to develop the fuels and materials for the prototype reactor Monju and future fast reactors. Accumulated data have been registered with OECD/NEA database with expectation that these data will be widely used. Joyo is presently temporary shutdown because of periodical inspection including in-vessel inspection and repair. After restart, Joyo will play a key role for a wide variety of science and technology fields as fast neutron irradiation bed.
Okawachi, Yasushi; Maeda, Shigetaka; Ito, Chikara; Kawahara, Hirotaka; Aoyama, Takafumi; Ishida, Koichi
JAEA-Technology 2009-047, 130 Pages, 2009/09
This report summarizes the contents about "Reactor physics and plant dynamics experiments using the Joyo simulator" which is one of the training themes. Training is performed using the full scope nuclear reactor simulator for Joyo operation training. While pushing from starting of a nuclear reactor in each experiment of criticality, a control rod proofreading examination, measurement of the temperature of a nuclear reactor, or the reactivity coefficient accompanying output change, feedback reactivity measurement of a fast reactor, etc. and understanding self-regulating characteristics peculiar to a nuclear reactor, the operation of a nuclear reactor can be experienced.