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Yokoyama, Keisuke; Uwaba, Tomoyuki
Journal of Nuclear Science and Technology, 60(10), p.1219 - 1227, 2023/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Nagase, Fumihisa
Annals of Nuclear Energy, 171, p.109052_1 - 109052_8, 2022/06
Times Cited Count：2 Percentile：50.96(Nuclear Science & Technology)The fracture threshold of the fuel decreases if the oxidized Zr alloy cladding is strongly constrained by the spacer grid during quenching in a loss-of-coolant accident. Therefore, the estimation of realistic levels of the axial constraint has been a subject of significant interest on fuel safety. In this study, a test assembly consisting of a PWR-type simulated fuel segment and a 3
3 grid piece was heated in steam, cooled, and quenched, and the axial constraint force on the fuel segment was measured. The constraint force of the Zircaloy grid gradually decreased with temperature. Once the Zircaloy grid was heated to 
1060 K, the reduced constraint force had difficulty recovering, and thus the maximum constraint force during cooling and quenching was 
10 N. The constraint force was clearly reduced at 1070 K during the tests with the Inconel grid. However, the reduced constraint force partially recovered during cooling. As a result, the maximum constraint force during cooling and quenching was 20 to 50 N for the Inconel grid. In conclusion, oxidation, ballooning, rupture, or eutectic formation would not generally cause an extremely strong constraint, as predicted by previous studies, at the grid position.
Uwaba, Tomoyuki; Yokoyama, Keisuke; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*; Pelletier, M.*
Nuclear Engineering and Design, 359, p.110448_1 - 110448_7, 2020/04
Times Cited Count：1 Percentile：11.8(Nuclear Science & Technology)Coupled computer code analyses of irradiation performance of axially heterogeneous mixed oxide (MOX) fuel elements with high burnup in a fast reactor were conducted. Post-irradiation experiments revealed local concentration of Cs near the interfaces between MOX fuel and blanket columns including the internal blanket of the fuel elements as well as an increase in their cladding diameters. The analyses indicated that the local Cs concentration occurred as a result of Cs axial migration from the MOX fuels toward the blanket pellets near the interfaces. Swelling of the blanket pellets induced by the formation of low-density Cs-U-O compound was not sufficient to cause pellet-to-cladding mechanical interaction (PCMI). The PCMI analyzed in the MOX fuel column regions was insignificant, and the cladding diameter increases were caused mainly by void swelling in cladding and irradiation creep due to fission gas pressure.
*; Kitada, Takanori*; Tagawa, Akihiro; *; Takeda, Toshikazu*
JNC TJ9400 2000-006, 272 Pages, 2000/02
JNC-TJ9400-2000-006.pdf:9.69MB
Investigation was made on the follwing three themes as a part of the improvement of reactor physics analysis method for FBR with various core concept. Part 1: Investigation of Error Estimation of Neutron Spectra in FBR and Suggestions to Improve the Accuracy. In order to improve the spectrum unfolding method used in fast experimental reactor JOYO, a trial was made to evaluate the error in the estimated neutron spectrum, cause by cause. And the evaluated errors were summed up to obtain the most probable and reasonable error as possible. The summed up error was found relatively small compared to the error caused by the uncertainty of cross section data: most of the error in the spectrum unfolding method can be attributed to the error in cross sections. It was also found that the error due to the fission spectrum causes a considerable error in the high energy neutron spectrum which is over several MeV. Part 2: Study on Reactor Physics Analysis Method for Gas-Cooled FBR. In gas-cooled FBR, the portion of coolant channels in core volume is larger than sodium-cooled FBR. This leads to strong neutron streaming effects. For sodium-cooled FBR, several methods were proposed to evaluate the neutron streaming effect, however, these methods can not be used directly to gas-cooled reactor because the direction dependent diffusion coefficient becomes infinitive along the direction pararel to the coolant chammel. In this study, a new method is proposed to evaluate the neutron streaming effect, based on the method taking the axial buckling into consideration, which method was originally proposed by K
hler. Part 3: Study on Reactor Physics Analysis Method for Water-Cooled FBR An investigation was made on low-moderated water-cooled FBR, on the point that the ordinary used analysis method for FBR may give considerable difference in results in such core. In light water reactors, it is well known that the space dependence of self-shielding effect of heavy nuclides are considerably ...
Takemura, Morio*
PNC TJ9055 97-001, 112 Pages, 1997/03
With use of a standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the Axial Shield Experiment (homogeneous and central blockage type shield configurations with B
C or stainless steel shield material) were performed. The results were compared with those obtained by the same analysis method and input data using JSDJ2 library that had been applied consistently to the JASPER experiment analyses. In general, the results with JSSTDL analyses are higher than those by JSDJ2 as were found in analyses in last year for the Radial Shield Attenuation Experiment and the Special Materials Experiment. Consideration was made on the discrepancies between JSSTDL and JSDJ2 analysis results of the Axial Shield Experiment and also those of the sodium configulation in the Radial Shield Attenuation Experiment. The former was done by exchange of macro cross section of each region, and the latter forcused on sodium cross section was done with use of cross section sensitivity analysis method. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments, that were selected in previous study in last year, were continued and new data were added into the computer disk holding previous ones.
Sakurai, Kiyoshi; Arakawa, Takuya*; Yamamoto, Toshihiro; Naito, Yoshitaka
JAERI-Research 96-045, 31 Pages, 1996/08
JAERI-Research-96-045.pdf:0.87MB
no abstracts in English
*; *; *
PNC TJ9614 94-001, 59 Pages, 1994/03
Crossflow of a two-phase mixture between vertical subchannels is subdivided into three components in the literature; turbulent mixing, void drift and diversion crossflow. Of these, turbulent mixing alone occurs in an equiliblium flow, in which flow rates of both phases in each subchannel do not change in the axial direction. In a general non-equilibrium flow, however, all three components occur simultaneously. In this report, effect of pressure differential between subchannels on flow redistribution process along the channel axis has been studied experimentally. In the experiment, a multiple channel, consisting of two identical circular subchannels of 16 mm I.D., were used as a test channel. And, air and water were introduced unevenly into the two subchannels at the inlet to get several non-equilibrium flows with and without the pressure differential between subchannels. For each flow, we have obtained the axial distributions data of pressure differential between the subchannels, the air and water flow rates, the void fractions, and the tracer concentrations for both phases when gas and liquid tracers were injected into one of the two subchannels. From these experimental data, we have estimated lateral velocities of the air and water corresponding to each crossflow component, and analyzed the effect of the pressure differential on the lateral velocities.
Chatani, Keiji; ; ; ; *; *; *
PNC TN9410 92-076, 348 Pages, 1992/03
JASPER (Japanese American Shielding Program of Experimental Researches) is the cooperative research program between PNC and US-DOE using TSF (Tower Shielding Facility) in ORNL (Oak Ridge National Laboratory) as the experiment facility. This report summarizes the works in FY'1991 as follows; Planning the experiment configuration for JASPER Program, Analyses of the JASPER Program experiment, Analyses of the former TSF experiment, and Development of the methods for FBR shielding analyses. (1)Analyses of the JASPER Program Experiment In FY'1991 Axial shield Experiment data were mainly analyzed, and some of In-vessel Fuel Storage (IVS) Experiment data were also analyzed. The Fast Reactor Shielding Analysis System developed by PNC has been applied to the analyses for JASPER Program experiments. (Axial Shield Experiment Analysis) Axial Shield Experiment was conducted from August 1990 through December 1990 as part of a continuing series of eight experimennts planned for the JASPER Program. The experiment serves not only to provide data for the verification of analysis system in calculating the neutron streaming in each design, but also to provide a basis for determining the shielding effectiveness of stainless steel (SS) and boron carbide (BC). four types of experimental configuration were used. The conclusions of the analyses are as follows: (a)For the spectrum modifier which provides a spectrum of neutron representative of those incident on the axial shield for the FBR core, the two-dimensional calculation showed good agreement with the experimental data. It was confirmed that the two-dimensional calculation could estimate the experimental data with almost the same accuracy as in the analyses for the Radial shield Attenuation and the Fission Gas Plenum Experiments. (b)For the homogeneous mockups, the two-dimensional ealculation could give the good agreement with the experimental data. (c)For the central blockage type mockups, in which the coolant flows ...
; Okumura, Keisuke; ; *; *
JAERI-M 92-030, 68 Pages, 1992/03
no abstracts in English
Iijima, Susumu; Okajima, Shigeaki; Obu, Makoto; Osugi, Toshitaka; ; ; *
Journal of Nuclear Science and Technology, 26(2), p.221 - 230, 1989/02
no abstracts in English
Iijima, Susumu; Obu, Makoto; *; Ono, Akio; ; Okajima, Shigeaki
Nuclear Science and Engineering, 100, p.496 - 506, 1988/12
Times Cited Count：2 Percentile：31.31(Nuclear Science & Technology)no abstracts in English
Ono, Akio; ;
Journal of Nuclear Science and Technology, 25(1), p.32 - 44, 1988/01
no abstracts in English
*; *; Tsukimori, Kazuyuki*
PNC TN9410 87-073, 206 Pages, 1987/03
The development of piping bellows as one of the cost reduction measures for Large FBR plants has been conducted in the Power Reactor and Nuclear Fuel Development Corporation since 1984. Fatigue and creep-fatigue strength evaluation is an important issue in practical application of bellows to the primary piping systems of FBR. This is the first report of a series of the fatigue and creep-fatigue tests for 42 inches diameter bellows. Fatigue tests described in this report were conducted on type 316 Stainless steel U-shaped bellows with seven convolutions at room temperature and 600 
C. Through the present tests and analytical studies, the following conclusions were obtained. (1)Many of fatigue cracks distribute on the root surfaces near the both sides of bellows and they are almost intergranular fracture. (2)Fundamental characteristics of bellows in elastic region such as spring and strain response can be properly estimated by EJMA standards and FEM analysis. (3)Fatigue life of bellows can be properly estimated by material based fatigue life prediction using measured strains of bellows. (4)Maximum actual strain ranges can be reasonably estimated by using the elastic nominal strain range multiplied by the inelastic amplification factor per convolution (f
f
) and the factor (f
) caused by stiffness fluctuation between convolutions. The effectiveness of the practical evaluation of strain range and fatigue life of bellows has been validated through this study.
-Scanning Method; Ono, Akio; ; *
JAERI-M 86-191, 76 Pages, 1987/01
no abstracts in English
; ; *; ; ; *; ; Ono, Akio; ; *; et al.
JAERI-M 86-065, 126 Pages, 1986/03
no abstracts in English
; Ueda, Shuzo
Atsuryoku Gijutsu, 24(5), p.254 - 263, 1986/00
no abstracts in English
-scanning Method; ; Ono, Akio
JAERI-M 85-207, 42 Pages, 1985/12
no abstracts in English
; ; *; ; ; ; ; Ono, Akio; *; ; et al.
JAERI-M 85-045, 136 Pages, 1985/04
no abstracts in English
