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Journal Articles

Determination of fusion barrier distributions from quasielastic scattering cross sections towards superheavy nuclei synthesis

Tanaka, Taiki*; Narikiyo, Yoshihiro*; Morita, Kosuke*; Fujita, Kunihiro*; Kaji, Daiya*; Morimoto, Koji*; Yamaki, Sayaka*; Wakabayashi, Yasuo*; Tanaka, Kengo*; Takeyama, Mirei*; et al.

Journal of the Physical Society of Japan, 87(1), p.014201_1 - 014201_9, 2018/01

 Times Cited Count:10 Percentile:72.43(Physics, Multidisciplinary)

Excitation functions of quasielastic scattering cross sections for the $$^{48}$$Ca + $$^{208}$$Pb, $$^{50}$$Ti + $$^{208}$$Pb, and $$^{48}$$Ca + $$^{248}$$Cm reactions were successfully measured by using the gas-filled recoil-ion separator GARIS. Fusion barrier distributions were extracted from these data, and compared with the coupled-channels calculations. It was found that the peak energies of the barrier distributions for the $$^{48}$$Ca + $$^{208}$$Pb and $$^{50}$$Ti + $$^{208}$$Pb systems coincide with those of the 2n evaporation channel cross sections for the systems, while that of the $$^{48}$$Ca + $$^{248}$$Cm is located slightly below the 4n evaporation ones. This results provide us helpful information to predict the optimum beam energy to synthesize superheavy nuclei.

Journal Articles

Thermal transient test and strength evaluation of a tubesheet structure made of Mod.9Cr-1Mo steel, 1; Test model design and experimental results

Ando, Masanori; Hasebe, Shinichi; Kobayashi, Sumio; Kasahara, Naoto*; Toyoshi, Akira*; Omae, Takahiro*; Enuma, Yasuhiro*

Nuclear Engineering and Design, 275, p.408 - 421, 2014/08

AA2013-0395.pdf:2.65MB

 Times Cited Count:2 Percentile:19.81(Nuclear Science & Technology)

To clarify the failure mode of a semispherical tubesheet structure originally designed for SG in the JSFR, a cyclic thermal loading test was performed using a tubesheet model test structure. The tubesheet model made of Mod.9Cr-1Mo steel was subjected to 1,873 cycles of severe thermal transient loads using a large-scale sodium loop, in which sodium heated to 600$$^{circ}$$C and 250$$^{circ}$$C was flowed repeatedly with periods for each transient of 2 and 1 h, respectively. After the test, the test model was inspected by PT. Then, observation using a SEM and hardness testing were performed. A thermal-hydraulic analysis was also performed to validate the measured temperature history during the thermal transient. Through these examinations and evaluation with thermal-hydraulic analysis, the manner of failure in the tubesheet under cyclic thermal loading is discussed.

Journal Articles

Thermal transient test and strength evaluation of a tubesheet structure made of Mod.9Cr-1Mo steel, 2; Creep-fatigue strength evaluation

Ando, Masanori; Hasebe, Shinichi; Kobayashi, Sumio; Kasahara, Naoto*; Toyoshi, Akira*; Omae, Takahiro*; Enuma, Yasuhiro*

Nuclear Engineering and Design, 275, p.422 - 432, 2014/08

AA2013-0396.pdf:1.44MB

 Times Cited Count:8 Percentile:60.89(Nuclear Science & Technology)

In this study, the strength of a tubesheet test model simulating a semispherical tubesheet structure subjected to cyclic thermal transients was evaluated using the finite element analysis (FEA). A test model made of Mod.9Cr-1Mo steel was subjected to 1,873 cycles of severe thermal transient loading using a large-scale sodium loop, in which elevated-temperature sodium at 600$$^{circ}$$C and 250$$^{circ}$$C was flowed repeatedly and kept at the final temperature for 2 and 1 h, respectively. Heat transfer analysis and stress analysis were performed using the sodium temperature data measured during the test. Then, the elastic and inelastic stress analysis results were used to investigate the failure mechanism by creep-fatigue damage and evaluate the failure strength. The evaluation based on the results of inelastic analysis estimated the number of cycles to failure within a factor of 3.

Journal Articles

Creep-fatigue evaluation of a structural model made of Mod.9Cr-1Mo steel subjected to the cyclic thermal loading

Ando, Masanori; Hasebe, Shinichi; Kobayashi, Sumio; Kasahara, Naoto*; Toyoshi, Akira*; Omae, Takahiro*; Enuma, Yasuhiro*

Nihon Kikai Gakkai M&M 2013 Zairyo Rikigaku Kanfarensu Koen Rombunshu (CD-ROM), p.OS1510_1 - OS1510_3, 2013/10

To validate the failure mode and assess creep-fatigue damage evaluation, a thick cylinder test model made of Mod.9Cr-1Mo steel was subjected to 1,873 cycles of accelerated thermal transient loading using a large-scale sodium loop through which liquid sodium at 600$$^{circ}$$C and 250$$^{circ}$$C flowed repeatedly, with the period of each transient being 2 h and 1 h, respectively. After completion of the test, liquid penetrant testing, a surface observation and hardness testing were performed to characterize failure mode. Based on the finite element analysis, creep-fatigue life was evaluated by applying the JSME FRs code. The failure cycles evaluated by rules described in the JSME FRs code was shown to have a safety margin of greater than 300 times for this system.

Journal Articles

Thermal transient test and strength evaluation of a thick cylinder model made of Mod.9Cr-1Mo steel

Ando, Masanori; Hasebe, Shinichi; Kobayashi, Sumio; Kasahara, Naoto*; Toyoshi, Akira*; Omae, Takahiro*; Enuma, Yasuhiro*

Nuclear Engineering and Design, 255, p.296 - 309, 2013/02

 Times Cited Count:17 Percentile:83.75(Nuclear Science & Technology)

To verify the failure mode and assess creep-fatigue damage, a thick cylinder test model made of Mod.9Cr-1Mo steel was subjected to 1,873 cycles of accelerated thermal transient loading using a large-scale sodium loop through which liquid sodium at 600 $$^{circ}$$C and 250 $$^{circ}$$C flowed repeatedly, with the period of each transient being 2 h and 1 h, respectively. After completion of the test, the test model was inspected using liquid penetrant testing. Observations using a scanning electron microscope and hardness testing were then performed to characterize creep-fatigue damage in the structural model subjected to cyclic thermal transient loading in a sodium environment. Finite element analysis were performed to evaluate the relationship between creep-fatigue damage and the observed crack conditions.

Journal Articles

New result in the production and decay of an isotope, $$^{278}$$113 of the 113th element

Morita, Kosuke*; Morimoto, Koji*; Kaji, Daiya*; Haba, Hiromitsu*; Ozeki, Kazutaka*; Kudo, Yuki*; Sumita, Takayuki*; Wakabayashi, Yasuo*; Yoneda, Akira*; Tanaka, Kengo*; et al.

Journal of the Physical Society of Japan, 81(10), p.103201_1 - 103201_4, 2012/10

 Times Cited Count:142 Percentile:97.35(Physics, Multidisciplinary)

An isotope of the 113th element, $$^{278}$$113, was produced in a nuclear reaction with a $$^{70}$$Zn beam on a $$^{209}$$Bi target. We observed six consecutive $$alpha$$ decays following the implantation of a heavy particle in nearly the same position in the semiconductor detector, in extremely low background condition. The fifth and sixth decays are fully consistent with the sequential decays of $$^{262}$$Db and $$^{258}$$Lr both in decay energies and decay times. This indicates that the present decay chain consisted of $$^{278}$$113, $$^{274}$$Rg (Z = 111), $$^{270}$$Mt (Z = 109), $$^{266}$$Bh (Z = 107), $$^{262}$$Db (Z = 105), and $$^{258}$$Lr (Z = 103) with firm connections. This result, together with previously reported results from 2004 and 2007, conclusively leads the unambiguous production and identification of the isotope $$^{278}$$113, of the 113th element.

Journal Articles

Spectra thermal fatigue tests under frequency controlled fluid temperature variation; Superposed sinusoidal temperature fluctuations tests

Kawasaki, Nobuchika; Takasho, Hideki*; Kobayashi, Sumio; Hasebe, Shinichi; Kasahara, Naoto

Proceedings of 2008 ASME Pressure Vessels and Piping Division Conference (PVP 2008) (CD-ROM), 9 Pages, 2008/07

To clarify frequency-dependent attenuation effects of fluid temperature fluctuation on fatigue strength, thermal fatigue strength tests subjected to superposed sinusoidal temperature fluctuations were performed by the SPECTRA test facility. After these fatigue tests, cylindrical test pieces were cut away from the test loop, and cracks were observed on the inner surface of the test pieces. Fatigue lives at crack initiation positions were evaluated based on the test conditions. Adopting power spectrum density functions and frequency transfer functions, fatigue lives were predicted within a factor 3.

JAEA Reports

Evaluation of creep strength on SUS304 under off-normal over-heating

Kato, Shoichi; Hasebe, Shinichi; Yoshida, Eiichi

JAEA-Research 2007-091, 33 Pages, 2008/02

JAEA-Research-2007-091.pdf:7.22MB

Loss of heat removal system at reactor shutdown is one of the objects of the probabilistic safety assessment of fast breeder reactors. In this research, the creep rupture data of type 304 stainless steel in high temperature has been obtained for the purpose of evaluation of mechanical strength of the structural materials at the severe accident, and applicability of the creep rupture formula adopted as the material strength standard for proto type fast breeder reactors has been investigated.

Journal Articles

Spectra thermal fatigue tests under frequency controlled fluid temperature variation; Strength tests

Kawasaki, Nobuchika; Hasebe, Shinichi; Kobayashi, Sumio; Kasahara, Naoto

Proceedings of 2007 ASME Pressure Vessels and Piping Division Conference/8th International Conference on Creep and Fatigue at Elevated Temperatures (PVP 2007/CREEP-8) (CD-ROM), 8 Pages, 2007/07

Thermal fatigue strength tests subjected to sinusoidal fluid temperature waves were performed by the SPECTRA test facility, where frequencies were 0.05, 0.2, and 0.5Hz. Cracks were observed on the inner surface of cylindrical test pieces after testing. 0.05Hz's wave caused a greater number of and deeper cracks than 0.5Hz's wave under the same fluid temperature range and the same fatigue cycles. The crack initiation region of the 0.05Hz's wave was larger than for the 0.5Hz's wave. Estimated fatigue failure cycles based on the frequency transfer functions were compared with test results. Frequency-dependency in failure cycles was observed through these test results, and frequency transfer functions could estimate this dependency. The test results supported the fatigue damage evaluation method with frequency transfer functions.

Journal Articles

Spectra thermal fatigue tests under frequency controlled fluid temperature variation; Transient temperature measurement tests

Kawasaki, Nobuchika; Kubayashi, Sumio; Hasebe, Shinichi; Kasahara, Naoto

Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 8 Pages, 2006/07

The coolant leakage by thermal striping phenomenon should be prevented at nuclear power plants and a lot of efforts are made to develop its evaluation methods. The frequency transfer function method can explain temperature and stress response to fluid temperature history using transfer function models; therefore it is considered as an excellent evaluation method. To measure temperature response of structures to fluid and to confirm their frequency characteristics, transient temperature measurement tests were performed by JAEA. In the transient temperature measurement tests, three different frequencies of sinusoidal fluid temperature waves (0.05, 0.2, 0.5Hz) were controlled using frequency controlled thermal fatigue test equipment (SPECTRA) and temperature responses at inner and outer structural surfaces were measured along the test sections. Frequency effects on temperature attenuation during transfer process from fluid to structures were confirmed and the effective heat transfer function in frequency transfer function method was verified by transient temperature measurement test results.

JAEA Reports

None

Nakayama, Fusao*; Enokido, Yuji*; Yoshida, Eiichi; Matsumoto, Toshiyuki; Hasebe, Shinichi

JNC TN9420 2005-001, 115 Pages, 2005/03

JNC-TN9420-2005-001.pdf:9.24MB

None

JAEA Reports

Material damage evaluation of welded structures model under thermal transient load

Hasebe, Shinichi; Onizawa, Takashi

JNC TN9400 2005-033, 66 Pages, 2005/03

JNC-TN9400-2005-033.pdf:6.54MB

In order to achive a longer operating life of FBR plant, evaluatiion technology to measure chages in the atructural materials due to damage mechanisms that cause aging and determine the remaining life, is nscessary. This report describe the results od evaluation tests to determine the remaining life of base materials and the micro damages of weld metal, using welded vesel model that have been damaged by creep fatigue in a thermal trasient test. The results obtained are as fallow.(1)Remaining life confirmation test of base materials (2)Micro damage confirmation test of welded metals

JAEA Reports

Experimental Study on Properties of High Cycle Thermal Fatigue, 3; Results of sinusoidal temperature fluctuation test at 20 second cycle

Hasebe, Shinichi; Kobayashi, Sumio; Tanaka, Hiroshi*; lbaraki, Koichi*; Fukasaku, Hiroshi*

JNC TN9400 2004-034, 73 Pages, 2004/03

JNC-TN9400-2004-034.pdf:6.02MB

In a nuclear power plant, it is necessary to be attentive to fatigue fracture of the structural material caused by cyclic thermal stress due to the mixing of temperature different fluids. The purpose of this study is to obtain data to demonstrate high cycle thermal fatigue evaluation methods by applying the effects of the frequency of temperature fluctuation. A sinusoidal temperature fluctuation test of with a 20 second period was conducted using high cycle fatigue test equipment (SPECTRA). A SUS304 steel pipe was used as the test sample, at an average sodium temperature of 425 deg-C, fluctuation amplitude of 200deg-C and a sodium flow rate of 300 l/min in the test pipe. The results obtained are as follows: (1)valid strength data to verify evaluation methods could be obtained by applying a 20 second cycle temperature fluctuation to the test sample with SPECTRA. A Crack penetrated at about 157,150 cycles. (2)Numerous cracks in an axial direction were observed on the jnner surface of the test sample in the upper flow area. An air fatigue test demonstrated the difference in the strength of the test sample between axial direction and circumferential direction, revealing that cracks were distributed in an axial direction since anisotropic influences easily appear on the hjgh cycle side. (3)An approximated curve obtained by the common relation of crack and axial direction distance indicates that the boundary of a crack would be located about 430 mm downstream from the tapered end of the test sample with the upper now. (4)Crack occurring on the inner surface progressed to a depth of 1 to 2 mm in thecrystal grain, then progressed along the crystal grain boundary. Striations were formed on areas of the fracture surface in the grain, but were not found on the fracture surface of the grain boundary. Sinusoidal temperature fluctuation tests at the periods of 2,5,10,and 40 seconds are planned to confirm the influence of fluctuation frequency responsiveness on structural material

Journal Articles

Spectra thermal fatigue tests under frequency controlled fluid temperature change; Development of test equipment and preliminary tests

kasahara, Naoto; Hasebe, Shinichi; Kobayashi, Sumio; Ando, Masanori; Kawasaki, Nobuchika; Morita, Hiroshi*

ASME PVP-Vol.472, P. 2986, 2004/00

Focusing on the cover layer materials (as the Radon Barrier Materials), which could have the effect to restrain the radon from scattering into the air and the effect of the radiation shielding, we produced the radon barrier materials with crude bentonite on an experimental basis, using the rotary type comprehensive unit for grinding and mixing, through which we carried out the evaluation of the characteristics thereof.

JAEA Reports

High Temperature Strength Properties of Mod.9Cr-1Mo Steel Forging Weld Zone

Hasebe, Shinichi; Kato, Shoichi

JNC TN9400 2003-101, 46 Pages, 2003/12

JNC-TN9400-2003-101.pdf:3.01MB

In this report, creep, fatigue and the creep-fatigue properties of Mod.9Cr-1Mo steel Forging weld, representing conventional steels, was evaluated in order to obtain the needed comparative data for the evaluation of the high-temperature strength of 12Cr steel weld.

JAEA Reports

Creep properties and microstructure change of FBR grade type 316 stainless steel weld zone

Hasebe, Shinichi; Onizawa, Takashi; Kato, Shoichi

JNC TN9400 2003-019, 62 Pages, 2003/03

JNC-TN9400-2003-019.pdf:3.33MB

We conducted a long-term creep test of weld metal and welded joint made from what we consider optimal filler metals for more than 10000 hours at 550$$^{circ}$$C, in order to evaluate their long-term properties at high temperatures, and select the appropriate filler metals for FBR grade type 316 stainless steel. We also conducted an evaluation of the long-term high temperature strength of the weld metals by observing changes in the microstructures that was subjected to material deterioration. The results obtained are as fo1lows. (1)Weld metal and welded joint made from 316 type and a specific optimal material for 16-8-2 type showed better creep properties than current materials. Especially for 16-8-2 type, the quality improved so much that predominance microstructure stability in the region of long-term. (2)We clearly showed that when $$delta$$-ferrite phase decomposed by long time heating. the Laves phase, $$sigma$$ phase and austenite phase were precipitated and the remaining $$delta$$ -ferrite phase was changed to an $$alpha$$-ferrite phase (Cr≒12%, Ni≒2%) as it became a low-alloy and reached equilibrium. (3)The long-term creep strength of the 316 type weld metal tends to decrease as $$sigma$$ phase separation increases due to a high Cr concentration in $$delta$$-ferrite phase. On the other hand, we confirmed that 16-8-2 type weld metal could maintain long-term creep strength almost as high as the base metal, because there was little separation of inter-metallic compounds due to its low concentration of Cr. (4)We found that change in the microstructure can be easily captured by analyzing the composition of the remaining $$delta$$-ferrite phase. This is an effective method to evaluate long-term high temperature strength.

JAEA Reports

Experimental study on properties of high cycle thermal fatigue; Outline and test plan of high cycle fatigue test equipment on sodium

Hasebe, Shinichi; Kobayashi, Sumio; Ando, Masanori; Kasahara, Naoto

JNC TN9400 2003-004, 110 Pages, 2003/01

JNC-TN9400-2003-004.pdf:4.23MB

At a nuclear power plant, where fluids of high and low temperatures flow into each other, it is necessary to prevent structural failure damage caused by high cycle thermal fatigue (thermal striping phenomenon). High cycle fatigue test equipment on thermal can be develop by modifying the thermal transient test facility for structure (TTS) in order to clarify the effect of temperature fluctuation induced by the thermal striping phenomenon on crack initiation and their propagation behavior. The test equipment has the following characteristic. (1)Fluid is controlled by a circulation pump, and by continuously changing the flow quantity ratio of high and low temperature Sodium, sinusoidal temperature fluctuations at various period of the test samples can be taken. (2)Mixing is done by the jet flow mix, thus it can generate axisymmetric temperature fluctuations by accelerating the mixing process of high and low temperature Sodium. (3)It can also control the temperature fluctuation, in which short and long term changes are superimposed. (4)Because the test sample cylinder is hollow, analysis of thermal stress and data from crack initiation to crack propagation can easily be obtained. Sinusoidal temperature fluctuations, random temperature fluctuations, and strength testing of the weld zone by test samples made of stainless steel are planned in the next stage.

JAEA Reports

Creep-fatigue life prediction method of 316FR welded joints, 1; Prediction based on macroscopic stress and strain concentration

Asayama, Tai; Hasebe, Shinichi

PNC TN9410 96-015, 39 Pages, 1996/01

PNC-TN9410-96-015.pdf:1.32MB

Form the viewpoint of cost reduction, it is necessary to avoid using expensive forged rings by adapting weldments in the parts where relatively large creep-fatigue damage is expected such as the part of pressure vessel to which the sodium surface contacts. Therefore the authors have been developing a creep-fatigue life prediction method for weldments. This report presents the newly developed prediction method ('3-element model') which describes the creep-fatigue life as well as the location of failure for 316FR weldments which is a candidate material for large scale FBRs. It was shown that the model satisfactory predicts the location of failure and that the prediction of life is possible within an accuracy of factor of 3 with a tendency of slight unconservatism for the results of mechanical fatigue/creep-fatigue tests with small and large specimens as well as for the thermal fatigue/creep-fatigue tests. Improvement of the accuracy of prediction is considered to be achieved by using more accurate dynamic stress-strain curve of weld metal and HAZ at lower strain range and more accurate fatigue strength of weld metal at lower strain range which are the base of the prediction. Thus fatigue tests of weld metal and HAZ at lower strain ranges are needed.

JAEA Reports

Material strength standard of FBR grade type 316 stainless steel(Draft)

Watashi, Katsumi; Aoto, Kazumi; Aoki, M; Komine, Ryuji; Ito, Takushi; Hasebe, Shinichi; Kato, Shoichi; Koi, Mamoru; Wada, Yusaku

PNC TN9410 93-142, 120 Pages, 1993/06

PNC-TN9410-93-142.pdf:6.08MB

Much progress has been made in improving established creep properties of Type 316 stainless steel and to develop a new structural material named "FBR Grade Type 316 Stainless Steel", 316FR, with superior creep properties. This report includes a draft of Material Strength Standard of 316FR and its interpretation on the basis of the major result of research and development conducted so far. The draft includes identical items described in the "Standards for the Strength of Materials" for Monju, and was carefully prepared to have an identical style for convenience in design evaluation. Creep damage evaluation diagrams, which are depicted in the "Structural Design Guide for Class 1 Components of Prototype Fast Breeder Reactor for Elevated Temperature Service" (ETSDG) for individual materials, are also included in this report.

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