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Journal Articles

Proposal of simulation material test technique for clarifying the structure failure mechanisms under excessive seismic loads

Hashidate, Ryuta; Kato, Shoichi; Onizawa, Takashi; Wakai, Takashi; Kasahara, Naoto*

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 9 Pages, 2021/07

It is very essential to clarify the structure failure mechanisms under excessive seismic loads. However, structural tests using actual structural materials are very difficult and expensive. Therefore, we have proposed the structure test approach using lead alloys in order to simulate the structure failure mechanisms under the excessive seismic loads. In this study, we conducted material tests using lead alloy and verified the effectiveness of the simulated material tests. Moreover, we formulated inelastic constitutive equations (best fit fatigue curve equation and cyclic stress range - strain range relationship equation) of lead alloy based on the results of a series of material tests. Nonlinear numerical analyses, e.g. finite element analyses, can be performed using the proposed equations. A series of simulation material test technique enables structural tests and analyses using lead alloy to simulate the structure failure phenomena under excessive seismic loads.

Journal Articles

Proposal of simulation materials test technique and their constitutive equations for structural tests and analyses simulating severe accident conditions

Hashidate, Ryuta; Kato, Shoichi; Onizawa, Takashi; Wakai, Takashi; Kasahara, Naoto*

Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 9 Pages, 2020/08

Although it is very essential to clarify how the structure collapses under the severe accident conditions, the failure mechanisms in excessive high temperatures are not clarified. However, it is very difficult and expensive to perform structural tests using actual structural materials. Therefore, we propose to use lead alloys instead of actual structural materials. For demonstration of analogy between the failure mechanisms of lead alloys structure at low temperature and those of the actual structures at high temperature, numerical analyses are required. Although the authors proposed inelastic constitutive equations for numerical analyses in 2019, the equations could not successfully express because of large variations observed in the material tests of the lead alloy. In this study, we propose the improved inelastic constitutive equations of the lead alloy on the basis of the material test results used by aged alloy which can stabilized the material characteristic.

Journal Articles

A Proposal of inelastic constitutive equations of lead alloys used for structural tests simulating severe accident conditions

Hashidate, Ryuta; Onizawa, Takashi; Wakai, Takashi; Kasahara, Naoto*

Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 10 Pages, 2019/07

Under the severe accident conditions, structural materials of nuclear power plants are subjected to excessive high temperature. Although it is very essential to clarify how the structure collapses under the severe accident conditions, the failure mechanisms in such high temperatures are not clarified. However, it is very difficult and expensive to perform structural tests using actual structural materials. Therefore, we propose to use lead alloys instead of actual structural materials. Because the strength of lead alloys is much poorer than that of the actual structural materials, failure can be observed at low temperature and by small load. For demonstration of analogy between the failure mechanisms of lead alloys structure at low temperature and those of the actual structures at high temperature, numerical analyses are required. So, we confirm the material characteristics of lead alloys and develop inelastic constitutive equations of lead alloy required for finite element analyses.

Journal Articles

Selector-valve failed fuel detection and location system for Japan Sodium-cooled Fast Reactor

Aizawa, Kosuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto*

Nuclear Technology, 189(2), p.111 - 121, 2015/02

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Selector-valve mechanism is adopted in the design of JSFR for its failed-fuel detection and location (FFDL) system. JSFR has only two FFDL units for 562 core fuel subassemblies to reduce construction cost by decreasing the reactor vessel diameter. Consequently, one SV-FFDL unit must handle about 300 subassemblies. In addition, JSFR adopts an upper internal structure (UIS) with a slit above the core. Sampling performance for the subassemblies under the UIS slit has been evaluated to be lower than those under the normal UIS position in the previous water experiments and numerical simulation. In this paper, the outline of FFDL system is shown, which can be applied to so large number of fuel subassemblies in a compact reactor vessel. Detection capability of the FFDL system was studied to achieve the design conditions. Operation modes and procedure of the FFDL system also investigated.

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

Development of constitutive models for fast reactor design

Tsukimori, Kazuyuki; Iwata, Koji*; Kawasaki, Nobuchika*; Okajima, Satoshi; Yada, Hiroki; Kasahara, Naoto*

Nuclear Engineering and Design, 269, p.23 - 32, 2014/04

 Times Cited Count:1 Percentile:10.7(Nuclear Science & Technology)

R&D to enable a practical fast breeder reactor plant is proceeding in Japan, which is called "FaCT" (Fast reactor Cycle Technology development). One of the key issues of R&D is to realize a reasonably compact reactor vessel by eliminating the wall protection equipment which is installed inside the vessel in order to reduce thermal loading in the conventional design. Most important concern is the amount of the inelastic strain of the vessel accumulated around the liquid sodium surface which moves upward and downward cyclically with start-up and shut-down. The aim of this study is to develop rational constitutive models that enable prediction of this kind of complex inelastic behaviors precisely and to prepare the design guide based on inelastic analysis. We developed a high accuracy plasticity model and a simplified plasticity model and valuated them by organized experiments.

Journal Articles

Stress mitigation design of a tubesheet by considering the thermal stress inducement mechanism

Ando, Masanori; Takasho, Hideki*; Kawasaki, Nobuchika; Kasahara, Naoto*

Journal of Pressure Vessel Technology, 135(6), p.061207_1 - 061207_10, 2013/12

 Times Cited Count:5 Percentile:31.84(Engineering, Mechanical)

The stress generation mechanism of a tubesheet was revealed through finite element analysis. Semi-spherical tubesheet models were investigated for the first survey of the thermal stress mechanism. The calculated results of the semi-spherical tubesheet model indicated an extensive peak stress around the outermost hole. The recognized thermal stress mechanism of a semi-spherical tubesheet is summarized, and on the basis of the stress generation mechanism, we proposed a stress-mitigated tubesheet, a center-flattened spherical tubesheet (CFST), as an improved configuration. The stress generation mechanism of the CFST was also desicibed.

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

Under-sodium endurance experiment of selector valve in failed-fuel detection and location system of JSFR

Aizawa, Kosuke; Fujita, Kaoru; Hirata, Shingo*; Kasahara, Naoto*

Nuclear Technology, 183(1), p.1 - 12, 2013/07

In the design of Japan Sodium-cooled Fast Reactor (JSFR), a selector-valve mechanism is adopted for its failed-fuel detection and location (FFDL) system. Since JSFR has only two FFDL units for about 600 fuel subassemblies, one FFDL unit must handle much larger number of subassemblies than in previous designs. In addition, during long plant life of 60 years, the wear length of the selector-valve will become longer than those of past reactors. Therefore, the endurance of the selector-valve becomes important. To demonstrate the manufacturability and endurance of the selector-valve, a full size mock-up was manufactured, and an endurance experiment of the mock-up model under high-temperature sodium were conducted. The cross-section observation, hardness measurement, and chemical assay results after the endurance experiment showed that the coating layer on the sliding surface still remains. Thus, the endurance of the JSFR selector-valve was demonstrated.

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

Evaluation on double-wall-tube residual stress distribution of sodium-heated steam generator by neutron diffraction and numerical analysis

Kisohara, Naoyuki; Suzuki, Hiroshi; Akita, Koichi; Kasahara, Naoto*

Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.621 - 630, 2012/06

Journal Articles

Water experiment and numerical simulation on failed fuel detection and location system of Japan sodium-cooled fast reactor (JSFR)

Aizawa, Kosuke; Oshima, Jun*; Kamide, Hideki; Kasahara, Naoto

Journal of Nuclear Science and Technology, 49(1), p.47 - 60, 2012/01

 Times Cited Count:2 Percentile:20.56(Nuclear Science & Technology)

A selector-valve type failed fuel detection and location (FFDL) system is applied to the JSFR design that has an upper internal structure (UIS) with a slit above the core and several sampling nozzles for the FFDL are set in the UIS around the slit to detect the fission product (FP) from the subassemblies below the slit. Therefore, mixing process in the UIS should be known and appropriate arrangement of the sampling nozzles in the UIS is needed. A water experiment using a 1/5-scle model was carried out. Experimental results showed that the sampling nozzles set in the UIS detected the FP simulant concentration within the criteria of FFDL signal detection. In addition, identification of the failed fuel subassembly under the UIS slit was achieved by means of comparing concentration profiles in the UIS. A numerical simulation using a CFD code was carried out. The simulation results showed that the simulation predicts the FP concentration distributions.

Journal Articles

Experimental and analytical study of failed fuel detection and location system in JSFR

Aizawa, Kosuke; Oshima, Jun*; Kamide, Hideki; Kasahara, Naoto

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00

JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product or delayed neutron. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. To overcome above diffculties, we have developed the sampling method for indentifying the failed fuel subassemblies located under the slit by numerical simulations and water experiments.

Journal Articles

Limitation study for cyclic hardening recovery for 316FR stainless steel derived from long-term holding with elevated temperature

Okajima, Satoshi; Kawasaki, Nobuchika*; Fukahori, Takuya*; Kikuchi, Koichi*; Kasahara, Naoto

Dai-49-Kai Koon Kyodo Shimpojiumu Koen Rombunshu, p.85 - 89, 2011/11

no abstracts in English

Journal Articles

Development of constitutive models for fast reactor design

Tsukimori, Kazuyuki; Iwata, Koji*; Kawasaki, Nobuchika*; Kasahara, Naoto*

Transactions of 21st International Conference on Structural Mechanics in Reactor Technology (SMiRT-21) (CD-ROM), 8 Pages, 2011/11

Journal Articles

The Creep-fatigue evaluation method for intermediate hold conditions; Proposal and validation

Okajima, Satoshi; Kawasaki, Nobuchika*; Kato, Shoichi; Kasahara, Naoto

Proceedings of 2011 ASME Pressure Vessels and Piping Conference (PVP 2011) (CD-ROM), 8 Pages, 2011/07

Journal Articles

Detection capability and operation patterns of a selector-valve failed-fuel detection and location system for large sodium-cooled reactors

Aizawa, Kosuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.605 - 613, 2011/05

A conceptual design study of an advanced large-sized (1,500 MWe class) sodium-cooled fast reactor, JSFR, is in progress in the FaCT project in Japan. JSFR has adopted a selector-valve mechanism for a failed-fuel detection and location (FFDL) system. The selector-valve FFDL system identifies a failed fuel subassembly by sampling sodium from each fuel subassembly outlet and detecting fission product gas or delayed neutron precursors of fission products. One of the technologies which JSFR has adopted is an upper internal structure (UIS) with a radial slit. Because sampling nozzles cannot be set in the UIS slit, several sampling nozzles are installed around the slit so as to sample sodium from the failed fuel subassemblies under the UIS slit. In this study, a signal and noise detected by the delayed neutron detector have been calculated. On the basis of these results, appropriate operation patterns of the selector-valve FFDL system for JSFR have been constructed.

Journal Articles

Development of failed fuel detection and location system in sodium-cooled large reactors; Sampling method of failed fuels under the slit

Aizawa, Kosuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

Nihon Kikai Gakkai Rombunshu, B, 77(776), p.982 - 986, 2011/04

A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the "Fast Reactor Cycle Technology Development (FaCT)" project in Japan. JSFR adopts a Selector-Valve mechanism for a failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, appropriate sampling method for indentifying under-slit failed fuel subassemblies has been developed by water experiments.

Journal Articles

Investigation on slit jet through upper internal structure (UIS) in highly compact vessel of sodium-cooled fast reactor

Kamide, Hideki; Aizawa, Kosuke; Oshima, Jun*; Nakayama, Okatsu*; Kasahara, Naoto

Journal of Nuclear Science and Technology, 47(9), p.810 - 819, 2010/09

 Times Cited Count:3 Percentile:26.36(Nuclear Science & Technology)

Development of advanced loop type sodium cooled fast reactor is under going. An upper internal structure (UIS) has a radial slit to reduce the reactor vessel diameter. This UIS slit allows a high velocity from the core fuel subassemblies and influences the gas entrainment in the reactor vessel and also the delayed neutron precursor sampling for a failed fuel detection and location system. Then flow visualization and velocity measurements were carried out in an 1/10 scale water test model. The velocity measurement using particle image velocimetry showed that velocity in the slit region was accelerated at the heights of the UIS horizontal plates and kept higher value at the middle height of the upper plenum. Numerical simulation using a commercial CFD code was also carried out for this complex geometry of UIS to know adequate simulation method. The comparisons of velocity profiles in the UIS between the experiment and analysis showed good agreements.

227 (Records 1-20 displayed on this page)