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Journal Articles

Study on fracture behaviour of through-wall cracked elbow under displacement control load

Machida, Hideo*; Koizumi, Yu*; Wakai, Takashi; Takahashi, Koji*

Nippon Kikai Gakkai M&M 2019 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.OS1307_1 - OS1307_5, 2019/11

This paper describes the fracture test and fracture analysis of a pipe under displacement control load. In order to grasp the fracture behavior of the circumferential through-wall cracked pipe, which is important in evaluating the feasibility of leak before break (LBB) in sodium cooled reactor piping, a fracture test in case of a circumferential throughwall crack in the weld line between an elbow and a straight pipe was carried out. From this test, it was found that no pipe fracture occurs in the displacement control loading condition even if a large circumferential through-wall crack (180$$^{circ}$$) was assumed. The fracture analysis of the pipe was carried out using Gurson's parameters set based on the tensile test results of the tested pipe material. The analytic results agree well with the test results, and it was found that it will be possible to predict the fracture behavior of sodium cooled reactor piping.

Journal Articles

Improvement of probabilistic fracture mechanics analysis code PASCAL-SP with regard to PWSCC

Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.

Journal of Nuclear Engineering and Radiation Science, 5(3), p.031505_1 - 031505_8, 2019/07

Probabilistic fracture mechanics (PFM) analysis is expected as a rational method for the structural integrity assessment because it can consider the uncertainties of various influence factors and can evaluate the quantitative value such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for the structural integrity assessment of piping welds in nuclear power plants. In the latest few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in the nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus the structural integrity assessment taking account of PWSCC has become important. In this paper, we improved PASCAL-SP for the assessment considering PWSCC by introducing the several analytical functions such as the evaluation models of crack initiation time, crack growth rate and probability of crack detection. By using improved PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were evaluated as numerical examples. We also evaluated the influence of a leak detection and a non-destructive examination on the failure probabilities. On the basis of the numerical results, we concluded that the improved PASCAL-SP is useful for evaluating the failure probability of pipe taking PWSCC into account.

Journal Articles

Development of FBG sensors by ultrashort pulse laser processing; Installation on high temperature pipeline and strain measurement

Nishimura, Akihiko

Dai-62-Kai Koha Senshingu Gijutsu Kenkyukai Koen Rombunshu, p.79 - 86, 2018/12

Thermal Energy Storage (TES) is important to stabilize increasing large amount of fluctuating renewable energy. For safety operation of TES, remote sensing by Fiber Bragg Grating (FBG) sensors is expected. FGB sensors were fabricated using precisely focused picosecond laser pulses. For the best use of heat resistant characteristic, we demonstrated to embed the FBG sensors in metal mold using colloidal silver adhesive. The FBG sensors were tested using a sodium circulation loop in JAEA Tsuruga site. Sodium was circulated with temperature of 500 degree. During emergency cooling, sudden shrinking of the loop was recorded. The application of FBG sensors to advanced remote monitoring for next generation TES plant is proposed.

Journal Articles

Prediction for plastic collapse stresses for pipes with inner and outer circumferential flaws

Hasegawa, Kunio; Li, Y.; Mares, V.*; Lacroix, V.*

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 5 Pages, 2018/07

Appendix C-5320 of ASME Code Section XI provides a formula of bending stress at the plastic collapse, where the formula is applicable for both inner and outer surface flaws. Authors considered the separated pipe mean radii at the flawed ligament and at the un-flawed ligament and formulas of plastic collapse stresses for each inner and outer flawed pipe were obtained. It is found that the collapse stress for inner flawed pipe is slightly higher than that calculated by Appendix C-5320 formula, and the collapse stress for outer flawed pipe is slightly lower than that by Appendix C-5320 formula. The collapse stresses derived from the three formulas are almost the same in most instances. For less common case where the flaw angle and depth are very large for thick wall pipes, the differences among the three collapse stresses become large.

Journal Articles

Improvement of probabilistic fracture mechanics analysis code PASCAL-SP with regard to primary water stress corrosion cracking

Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07

Recently, cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel based alloy welds in the primary piping of pressurized water reactors. Structural integrity assessments taking PWSCC into account have become important. Probabilistic fracture mechanics (PFM) is expected as one of rational methods for the assessments because it can account for uncertainty of the influencing factors and evaluate the failure probabilities of components. In JAEA, a PFM analysis code PASCAL-SP was developed to evaluate the failure probability of nuclear pipe. This paper details improvement of the PASCAL-SP to evaluate the failure probability taking PWSCC into account. As numerical examples, the failure probabilities for circumferential and axial cracks due to PWSCC are evaluated. Influence of inspection on failure probabilities are evaluated. As the results, we conclude that the improved PASCAL-SP is useful for evaluating the failure probability taking PWSCC into account.

Journal Articles

Development of picosecond laser writing for heat resistant FBG sensors and adhesion technique for high temperature industrial plants

Nishimura, Akihiko; Takenaka, Yusuke*

Sumato Purosesu Gakkai-Shi, 6(2), p.74 - 79, 2017/03

no abstracts in English

Journal Articles

Characteristics of flow field and pressure fluctuation in complex turbulent flow in the third elbow of a triple elbow piping with small curvature radius in three-dimensional layout

Ebara, Shinji*; Takamura, Hiroyuki*; Hashizume, Hidetoshi*; Yamano, Hidemasa

International Journal of Hydrogen Energy, 41(17), p.7139 - 7145, 2016/05

 Times Cited Count:3 Percentile:85.56(Chemistry, Physical)

In this study, a flow visualization and pressure measurement were conducted by using an experimental setup including test sections of 1/7-scale models of the cold-leg piping of Japan sodium-cooled reactor with high Reynolds number region up to about one million. Regarding the flow field, flow separation appeared in the intrados of the third elbow. However, the separation region was smaller than that observed in the first elbow in the direction normal to the mean flow and was larger in the streamwise direction. This can be considered because of the swirling flow generated downstream of the second elbow which flowed into the third elbow with a little reduction. From the pressure fluctuation test, it was found that prominent frequencies of the pressure fluctuation appeared at about 0.4 in Strouhal number, which corresponds to a nondimensional frequency, in the region from 0 D to 0.4 D downstream of the elbow outlet, where D is the diameter of the piping. And weak peaks of about 0.7 in Strouhal number were observed in the region far 0.75 D downstream of the outlet.

Journal Articles

Flow-induced vibration evaluation of primary hot-leg piping in advanced loop-type sodium-cooled fast reactor for demonstration

Yamano, Hidemasa; Xu, Y.*; Sago, Hiromi*; Hirota, Kazuo*; Baba, Takeo*

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1029 - 1038, 2016/04

This study conducted the flow-induced vibration evaluation of the primary hot-leg piping in the demonstration reactor design of advanced loop-type sodium-cooled fast reactor in order to confirm the integrity of the piping. Following the description of the primary hot-leg piping design and a design guideline of the flow-induced vibration evaluation, this paper describes mainly the flow-induced vibration evaluation and thereby the integrity assessment. In the fatigue evaluation for the flow-induced vibration, the pipe stresses considering the stress concentration factor and so on, at representative locations were less than the design fatigue limit. Therefore, this evaluation confirmed the integrity of the primary hot-leg piping in the demonstration reactor.

Journal Articles

Characteristics of flow field and pressure fluctuation in complex turbulent flow in the third elbow of a triple elbow piping with small curvature radius in three-dimensional layout

Ebara, Shinji*; Takamura, Hiroyuki*; Hashizume, Hidetoshi*; Yamano, Hidemasa

Proceedings of 17th International Conference on Emerging Nuclear Energy Systems (ICENES 2015) (CD-ROM), 6 Pages, 2015/10

In this study, a flow visualization and pressure measurement were conducted by using an experimental setup including test sections of 1/7-scale models of the cold-leg piping of Japan sodium-cooled reactor with high Reynolds number region up to about one million. Regarding the flow field, flow separation appeared in the intrados of the third elbow. However, the separation region was smaller than that observed in the first elbow in the direction normal to the mean flow and was larger in the streamwise direction. This can be considered because of the swirling flow generated downstream of the second elbow which flowed into the third elbow with a little reduction. From the pressure fluctuation test, it was found that prominent frequencies of the pressure fluctuation appeared at about 0.4 in Strouhal number, which corresponds to a nondimensional frequency, in the region from 0 D to 0.4 D downstream of the elbow outlet, where D is the diameter of the piping. And weak peaks of about 0.7 in Strouhal number were observed in the region far 0.75 D downstream of the outlet.

Journal Articles

New instrumentation using a heat resistant FBG sensor installed by laser cladding

Nishimura, Akihiko; Terada, Takaya; Takenaka, Yusuke*; Furuyama, Takehiro*; Shimomura, Takuya

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07

Since 2007, JAEA has been developing laser based technologies of structural health monitoring. The FBG sensor made by femtosecond laser processing will be the best candidate. To make the best use of the heat resistant characteristic, the FBG sensor was embedded in metal mold by laser cladding. A groove was processed to the surface of a SUS metal plate. We used a QCW laser to weld a filler wire on the plate. A series of weld beads perfectly formed a sealing clad on the groove. Though the FBG sensor was buried tightly, no degradation on the reflection spectrum was detected after the processing. The FBG sensor could detect the vibration of the plate caused by impact shocks and audio vibration. The reflection peak of the FBG sensor under laser cladding condition was shifted to be 6 nm. We demonstrated that the corresponded temperature derive from the reflection peak shift reached 600 degrees in heat shock experiments. The installation procedure of a FBG sensor using a portable laser cladding machine was described.

JAEA Reports

User's manuals of probabilistic fracture mechanics codes PASCAL-SC and PASCAL-EQ

Ito, Hiroto*; Onizawa, Kunio; Shibata, Katsuyuki*

JAERI-Data/Code 2005-007, 118 Pages, 2005/09

JAERI-Data-Code-2005-007.pdf:5.23MB

As a part of the aging and structual integrity research for LWR components, new PFM (Probabilistic Fracture Mechanics) codes PASCAL-SC and PASCAL-EQ have been developed. These codes evaluate the failure probability of an aged welded joint by Monte Carlo method. PASCAL-SC treats Stress Corrosion Cracking (SCC) in piping, while PASCAL-EQ takes fatigue crack growth by seismic load into account. The development of these codes has been aimed to improve the accuracy and reliability of analysis by introducing new analysis and methodologies and algorithms considering the recent development in the fracture machanics methodologies and computer performance. The crack growth by an irregular stress due to seismic load in detail is considered in these codes. They also involves recent stress intensity factors and fracture criteria. In addition, a user's friendly operation of a GUI (Graphical User Interface) which generates input data, supports calculations and plots results is introduced. This report provides the user's manual and theoretical background of these codes.

JAEA Reports

Report of Examination of the Samples from Primary Loop Recirculation Piping (K1-PLR) at Kashiwazaki-Kariwa Nuclear Power Station Unit-1 (Contract Research)

The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi

JAERI-Tech 2004-049, 44 Pages, 2004/06

JAERI-Tech-2004-049.pdf:7.21MB

At the Kashiwazaki-Kariwa Nuclear Power Station Unit-1, indications of cracks were identified in a weld joint portion of the primary loop recirculation piping. To investigate the cause of cracks, TEPCO conducted a material examination on the specimen including the cracks. The present investigation was carried out to ensure transparency of the examination by providing JAERI's own evaluation report as a third party organization. The following findings were made; (1) A crack was observed near the weld region. (2) Intergranular cracking was observed at almost whole fracture surface. (3) Transgranular cracking was observed at the crack opening region. Increases of hardness by cold work were observed and the crack was initiated near the region where hardness value showed the highest. (4) Content of Cr was very slightly depleted in the vicinity of grain boundary. Based on the above results with the presence of tensile residual stress near the crack generated by welding process and dissolved oxygen contents in cooling water, the observed cracks were concluded to be stress corrosion cracking.

JAEA Reports

Report of Examination of the Samples from Primary Loop Recirculation Piping (O1-PLR) at Onagawa Nuclear Power Station Unit-1 (Contract research)

The Working Team for Examination Operation of Samples from Primary Loop Recirculation Piping at Onag

JAERI-Tech 2004-003, 74 Pages, 2004/02

JAERI-Tech-2004-003.pdf:30.08MB

The present examination has been performed with the objective to provide technical basis for identifying causes of cracking through the examination of the samples, which was conducted at the post irradiation examinations facilities of JAERI, taken from the cracked region of the recirculation pipe at the Onagawa Nuclear Power Station Unit-1. The following findings were obtained from this examination result. (1) Cracks were observed near the weld region of inner surface of the pipe and depth of the crack was about 5 to 7mm. (2) Intergranular crackings were observed in the almost whole fracture surface. Partially transgranular cracking was observed at the surface layer with the depth of about 100$$mu$$m. Microstructure formed by cold work and increase of hardness were observed in these surface layers. Cracks initiated near the region where hardness value was the highest. Based on the examination results described above concerning presence of tensile residual stress by welding and relatively high dissolved oxygen contents in core coolant and so on, it is concluded that these cracks were initiated in the cold work layer of inner surface by stress corrosion cracking (SCC) and propagated along the grain boundaries.

JAEA Reports

Annual report on operation, utilization and technical development of Hot Laboratories; April 1 2002 to March 31 2003

Department of Hot Laboratories

JAERI-Review 2003-038, 106 Pages, 2003/12

JAERI-Review-2003-038.pdf:9.36MB

no abstracts in English

Journal Articles

Design of mercury cirulation system for J-SNS

Kinoshita, Hidetaka; Haga, Katsuhiro; Kogawa, Hiroyuki; Kaminaga, Masanori; Hino, Ryutaro

Proceedings of ICANS-XVI, Volume 3, p.1305 - 1314, 2003/07

The JAERI and the KEK are promoting a plan to construct the spallation neutron source at the Tokai Research Establishment, JAERI, under J-PARC project. A mercury circulation system has been designed so as to supply mercury to the target stably. Conceptual design is almost finished. But, it was necessary to confirm a mercury pump performance, and more, to investigate erosion rate under the mercury flow as well as an amount of mercury remained on the surface after drain. The mercury pump performance was tested under the mercury flow conditions by using an experimental gear pump, which had almost the same structure as a practical mercury pump to be expected, and the erosion rates in a mercury pipeline as were investigated. The discharged flow rates of the gear pump increased linearly with the rotation speed. Erosion rates obtained under the mercury velocity less than 1.6 m/s was found to be so small. For the amount of remaining mercury on the pipeline, radioactivity of this remaining mercury volume was found to be three-order lower than that of the target casing.

Journal Articles

Helium production due to neutron streaming through small circular ducts in a fusion reactor blanket by analytical fitting from Monte Carlo calculation results

Sato, Satoshi; Nakamura, Takashi*; Nishitani, Takeo

Fusion Science and Technology, 43(4), p.559 - 568, 2003/06

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Inhibition of sensitization in reactor pipe materials by grain boundary structure control, JAERI's nuclear research promotion program, H11-023 (Contract research)

Kokawa, Hiroyuki*; Shimada, Masayuki*; Wang, Z.*; Sato, Yutaka*; Sato, Yoshihiro*; Kiuchi, Kiyoshi

JAERI-Tech 2003-014, 22 Pages, 2003/03

JAERI-Tech-2003-014.pdf:1.68MB

no abstracts in English

JAEA Reports

Annual report on operation, utilization and technical development of Hot Laboratories; April 1, 2001 to March 31, 2002

Department of Hot Laboratories

JAERI-Review 2002-039, 106 Pages, 2003/01

JAERI-Review-2002-039.pdf:9.46MB

no abstracts in English

Journal Articles

Development of laser cutting/welding system for remote maintenance of ITER manifold

Yamaoka, Hiroto*; Tsuchiya, Kazuyuki*; Awano, Toshihiko*; Oka, Kiyoshi

Ishikawajima Harima Giho, 42(5), p.260 - 264, 2002/09

no abstracts in English

JAEA Reports

Mercury flow experiments, 4; Measurements of erosion rate caused by mercury flow

Kinoshita, Hidetaka; Kaminaga, Masanori; Haga, Katsuhiro; Hino, Ryutaro

JAERI-Tech 2002-052, 28 Pages, 2002/06

JAERI-Tech-2002-052.pdf:7.59MB

Since the Neutron Scattering Facility will be using mercury as the target material and contain radioactive products, it is necessary to estimate reliability of instruments in a system. The system would be damaged by erosion. An erosion test section and coupons were installed in the mercury loop, and their thickness was measured. As a result, the erosion is about 3$$mu$$m in 1000 hours under 0.7m/s condition. The wall thickness decrease during facility lifetime of 30 years is estimated to be less than 0.5mm. Therfore, the effect of erosion on component strength is extremely small. Moreover, a measurement of residual mercury on the piping surface was carried out. As a result, 19g/m$$^{2}$$ was obtained. Thus, estimation of residual mercury for 150A-sch80 piping is 8.5g/m, and for the mercury target is about 40g. As for the target, radioactivity of the residual mercury is 1.2$$times$$10$$^{12}$$ Bq, which is extremely lower than that in the target casing of 1.0$$times$$10$$^{15}$$ Bq. Then, there is no influence for maintenance and storage of the spent mercury target.

241 (Records 1-20 displayed on this page)