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Journal articles on the topic 'Fracture mechanics. Materials Nuclear power plants'

Author: Grafiati

Published: 4 June 2021

Last updated: 15 February 2022

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1

Yoshida, K., M. Kojima, M. Iida, and I. Takahashi. "Fracture toughness of weld metals in steel piping for nuclear power plants." International Journal of Pressure Vessels and Piping 43, no. 1-3 (1990): 273–84. http://dx.doi.org/10.1016/0308-0161(90)90107-s.

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2

Timofeev, B. T., R. P. Vinogradov, S. P. Generalova, and T. A. Chernaenko. "Fracture-resistance evaluation for piping materials of BWR nuclear power plants of the RBMK type." International Journal of Pressure Vessels and Piping 52, no. 3 (1992): 303–11. http://dx.doi.org/10.1016/0308-0161(92)90088-w.

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3

Yagawa, G., Y. Ando, K. Ishihara, T. Iwadate, and Y. Tanaka. "Stable and Unstable Crack Growth of A508 Class 3 Plates Subjected to Combined Force of Thermal Shock and Tension." Journal of Pressure Vessel Technology 111, no. 3 (1989): 234–40. http://dx.doi.org/10.1115/1.3265669.

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An urgent problem for nuclear power plants is to assess the structural integrity of the reactor pressure vessel under pressurized thermal shock. In order to estimate crack behavior under combined force of thermal shock and tension simulating pressurized thermal shock, two series of experiments are demonstrated: one to study the effect of material deterioration due to neutron irradiation on the fracture behavior, and the other to study the effect of system compliance on fracture behavior. The test results are discussed with the three-dimensional elastic-plastic fracture parameters, J and Jˆ int

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4

Yoon, Ji-Hyun, Bong-Sang Lee, and Jun-Hwa Hong. "J-R fracture characteristics of ferritic steels for RPVs and RCS piping of nuclear power plants." Metals and Materials International 7, no. 5 (2001): 505–12. http://dx.doi.org/10.1007/bf03027094.

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5

Siegl, Jan, Petr Haušild, Adam Janča, Radim Kopřiva, and Miloš Kytka. "Characterisation of Mechanical Properties by Small Punch Test." Key Engineering Materials 606 (March 2014): 15–18. http://dx.doi.org/10.4028/www.scientific.net/kem.606.15.

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The specific desired properties for structures and components working in critical environments (e.g. different structure parts of power plants) require current information about degradation processes coming out in materials. Obtaining of this information by the help of the classical tests of mechanical properties (tensile test, Charpy test, fracture toughness test, creep test etc.) is very limited namely in the case of nuclear power plants pressure vessel. Hence, the new innovative techniques based on miniaturized specimens have been developed for evaluation of mechanical properties and their

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6

Naizabekov, Abdrakhman, Alexandr Arbuz, Sergey Lezhnev, Evgeniy Panin, and Marcin Knapinski. "Study of Technology for Ultrafine-Grained Materials for Usage as Materials in Nuclear Power." New Trends in Production Engineering 2, no. 2 (2019): 114–25. http://dx.doi.org/10.2478/ntpe-2019-0077.

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Abstract Nuclear power is associated with great environmental risks. In many cases, the problem of accidents of nuclear power plants is related to the use of materials that do not fully meet the following requirements: high corrosion resistance; high temperature resistance; creep resistance; fracture toughness; stability of structure and properties under irradiation. Therefore, studies aimed at finding materials that can withstand long-term loads at high temperatures, aggressive environment and gradual structural degradation under the influence of radiation are relevant. One of the structural

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7

Jundong, Lu, Jiang Xiaobin, Sun Ke, Liu Bin, Li Xinmin, and Ni Qinwen. "Stress Corrosion Cracking Behavior of TP 439 and 690 TT under Film-Forming Amine Environment." Scanning 2021 (June 8, 2021): 1–8. http://dx.doi.org/10.1155/2021/6668537.

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Film-forming amines have been widely used in thermal power plants for maintenance after shutdown, and there are more and more applications and researches in nuclear power secondary circuits for this purpose. However, in the direction of stress corrosion cracking, there is not much research on the influence of film-forming amines on metal materials. This article uses the high temperature slow strain rate test (SSRT) method to evaluate the influence of a commercial film-forming amine on the stress corrosion cracking behavior of two conventional island materials for PWR nuclear power plants. Thes

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8

Torop, V. M., M. D. Rabkina, O. O. Shtofel’, V. V. Usov, N. M. Shkatulyak, and O. S. Savchuk. "On the Causes of Fractures of Reinforcing Ropes of the Protective Shells of Power-Generating Units of Nuclear Power Plants." Materials Science 54, no. 2 (2018): 240–49. http://dx.doi.org/10.1007/s11003-018-0179-y.

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9

Kim, J. C., Sang Min Lee, Yoon Suk Chang, Jae Boong Choi, Young Jin Kim, and Young Hwan Choi. "Development of an Integrity Evaluation System for Steam Generator Tubes in a Nuclear Power Plant." Solid State Phenomena 120 (February 2007): 157–62. http://dx.doi.org/10.4028/www.scientific.net/ssp.120.157.

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Steam generators working in nuclear power plants convert water into steam from heat produced in the reactor core and each of them contains from 3,000 to 16,000 tubes. Since these tubes constitute one of primary barriers under radioactive and high pressure condition, the integrity should be maintained carefully during the operation. The objective of this research is to introduce an integrity evaluation system for steam generator tubes as a substitute of well-trained engineers or experts. For this purpose, a couplet examination has been carried out on the complicated evaluation procedure and an

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10

Wang, Wei Bing, He Xue, Fu Qiang Yang, and Kun Liu. "Effects of Grain Size on Crack Tip Mechanical Fields of Intergranular Cracking." Advanced Materials Research 1004-1005 (August 2014): 1147–51. http://dx.doi.org/10.4028/www.scientific.net/amr.1004-1005.1147.

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Intergranular stress corrosion cracking is one of important failure form of structural materials in nuclear power plants, and the initiation and development of crack at grain boundary are affected by the grain size of materials. The macroscopic model and mesoscopic model of crack propagation was established by using finite element method, and the effects of grain size on fracture parameters such as Mises stress, the maximum principal stress and equivalent plastic strain nearby crack tip were studied. The results indicate that the distribution of Mises stress and equivalent plastic strain are d

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11

Rieth, Michael, Dave Armstrong, Bernhard Dafferner, et al. "Tungsten as a Structural Divertor Material." Advances in Science and Technology 73 (October 2010): 11–21. http://dx.doi.org/10.4028/www.scientific.net/ast.73.11.

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Refractory materials, in particular tungsten base materials are considered as primary candidates for structural high heat load applications in future nuclear fusion power plants. Promising helium-cooled divertor design outlines make use of their high heat conductivity and strength. The upper operating temperature limit is mainly defined by the onset of recrystallization but also by loss of creep strength. The lower operating temperature range is restricted by the use of steel parts for the in- and outlets as well as for the back-bone. Therefore, the most critical issue of tungsten materials in

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12

Du, L. Y., G. Z. Wang, F. Z. Xuan, and S. T. Tu. "Effects of local mechanical and fracture properties on LBB behavior of a dissimilar metal welded joint in nuclear power plants." Nuclear Engineering and Design 265 (December 2013): 145–53. http://dx.doi.org/10.1016/j.nucengdes.2013.07.028.

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13

Xue, Fei, Fangjie Shi, Chuangju Zhang, et al. "The Microstructure and Mechanical and Corrosion Behaviors of Thermally Aged Z3CN20-09M Cast Stainless Steel for Primary Coolant Pipes of Nuclear Power Plants." Coatings 11, no. 8 (2021): 870. http://dx.doi.org/10.3390/coatings11080870.

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The effects of thermal aging time at 400 °C on the microstructure and mechanical and corrosion behaviors of Z3CN20.09M cast stainless steel were investigated; and the corresponding thermal aging mechanism was studied. It was revealed that the changes in mechanical properties after thermal aging were mainly caused by the iron-rich phase (α) and the chromium-rich phase (α’) produced by the amplitude-modulation decomposition of ferrite. A similar trend of thermoelectric potential during thermal aging was determined in relation to the Charpy impact energy. However, the corrosion resistance of Z3CN

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14

Lee, Tae-kyung, Seokmin Hong, Jongmin Kim, Min-Chul Kim, and Jae-il Jang. "Evaluation of Transition Temperature in Reactor Pressure Vessel Steels 6using the Fracture Energy Transition Curve from a Small Punch Test." Korean Journal of Metals and Materials 58, no. 8 (2020): 522–32. http://dx.doi.org/10.3365/kjmm.2020.58.8.522.

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The small punch (SP) test is one of the small specimen test techniques, and standardization of the SP test method for evaluating the mechanical properties of metallic materials is in progress. In this study, the impact transition temperature of reactor pressure vessel steels (RPV) in nuclear power plants was estimated using the draft standard SP test method. The SP fracture energy (ESP) and normalized SP fracture energy (ENSP) of the RPV steels were evaluated at various temperatures, and their transition curves were derived and compared to the transition curve in the Charpy V notch (CVN) test.

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15

Cizelj, L., B. Mavko, and P. Vencelj. "Reliability of Steam Generator Tubes With Axial Cracks." Journal of Pressure Vessel Technology 118, no. 4 (1996): 441–46. http://dx.doi.org/10.1115/1.2842211.

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An approach for estimating the failure probability of tubes containing through-wall axial cracks has already been proposed by the authors. It is based on probabilistic fracture mechanics and accounts for scatter in tube geometry and material properties, scatter in residual and operational stresses responsible for crack propagation, and characteristics of nondestructive examination and plugging procedures (e.g., detection probability, sizing accuracy, human errors). Results of preliminary tests demonstrated wide applicability of this approach and triggered some improvements. The additions to th

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16

Park, Dae Kyu, Seung Wan Woo, Il Sup Chung, Young Suck Chai, and Jae Do Kwon. "Evaluation of Fretting Fatigue Life for Steam Generator Tubes in Nuclear Power Plants." Key Engineering Materials 345-346 (August 2007): 243–46. http://dx.doi.org/10.4028/www.scientific.net/kem.345-346.243.

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Studies on the strength and fatigue life of machines and structures have been conducted in accordance with the development of modern industries. In particular, fine and repetitive cyclic damage occurring in contact regions has been known to have an impact on fretting fatigue fractures. INCONEL alloy 600, 690 and INCOLOY alloy 800 are iron-nickel-chromium alloy having excellent resistance to many corrosive aqueous media and high-temperature atmospheres. These alloy are used extensively in the nuclear power plants industry, the chemical industry, the heat-treating industry and the electronic ind

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17

Shinozaki, S. S., J. Hangas, K. R. Carduner, M. J. Rokosz, K. Suzuki, and N. Shinohara. "Correlation between microstructure and mechanical properties in silicon carbide with alumina addition." Journal of Materials Research 8, no. 7 (1993): 1635–43. http://dx.doi.org/10.1557/jmr.1993.1635.

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The microstructure of pressureless sintered silicon carbide (SiC) materials with alumina (Al2O3) addition was investigated using analytical electron microscopy and nuclear magnetic resonance. A sintered body with a density of higher than 99% theoretical was obtained with an addition of 5 wt.% Al2O3. The sintered body (SiC–Al2O3) has high strength, high fracture toughness, and high fatigue resistance. Its fracture toughness is approximately 5 MPa-m1/2, which is twice as high as that of pressureless sintered SiC materials with boron and carbon additions (SiC–B–C). The correlation between the mic

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18

Noel, R. "Life duration of PWR nuclear power plants." International Journal of Pressure Vessels and Piping 32, no. 5 (1988): 415–36. http://dx.doi.org/10.1016/0308-0161(88)90146-9.

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19

Faidy, C., S. Bhandari, and P. Jamet. "Leak-before-break in French nuclear power plants." International Journal of Pressure Vessels and Piping 43, no. 1-3 (1990): 151–63. http://dx.doi.org/10.1016/0308-0161(90)90098-3.

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20

Frolov, K. V., N. A. Makhutov, S. M. Kaplunov, V. A. Petushkov, L. V. Smirnov, and V. F. Ovchinnikov. "Vibration stability of main circulation pipelines in nuclear power plants." Strength of Materials 17, no. 10 (1985): 1337–46. http://dx.doi.org/10.1007/bf01534022.

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21

Kander, Ladislav, Petr Čížek, Šárka Hermanová, and Zdeněk Říha. "Structure and Mechanical Properties of Welded Joints for Nuclear Power Plants of Type MIR 1200." Materials Science Forum 891 (March 2017): 201–5. http://dx.doi.org/10.4028/www.scientific.net/msf.891.201.

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The paper deals with research, development and verification of production technology of selected welded joints for pressure vessels of primary circuits of nuclear power plants of type MIR 1200. Effect of various welding technology including simulation heat treatment on mechanical and fracture properties have been studied. Four type of homogenous 10GN2MFA – 10GN2MFA type of welded joints have been prepared for experimental programme. Conventional mechanical properties (tensile and impact test) as well as unconventional mechanical properties (fracture mechanics, low-cycle fatigue and stress corr

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22

Fridman, N. A., V. A. Strizhalo, A. F. Voitenko, I. N. Butkin, and M. P. Zemtsov. "Comprehensive examination of the technical state of the metal of fittings of power units of nuclear power plants." Strength of Materials 30, no. 2 (1998): 114–23. http://dx.doi.org/10.1007/bf02811276.

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23

Kilian, R., and A. Roth. "Corrosion behaviour of reactor coolant system materials in nuclear power plants." Materials and Corrosion 53, no. 10 (2002): 727–39. http://dx.doi.org/10.1002/1521-4176(200210)53:10<727::aid-maco727>3.0.co;2-#.

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24

Katsuyama, J., H. Itoh, Y. Li, K. Osakabe, K. Onizawa, and S. Yoshimura. "Benchmark analysis on probabilistic fracture mechanics analysis codes concerning fatigue crack growth in aged piping of nuclear power plants." International Journal of Pressure Vessels and Piping 117-118 (May 2014): 56–63. http://dx.doi.org/10.1016/j.ijpvp.2013.10.010.

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25

Gorynin, I. V., V. A. Ignatov, Yu I. Zvezdin, B. T. Timofeev, and V. M. Filatov. "Fracture resistance of welded thick-walled high-pressure vessels in power plants. Report No. 1. Statistical analysis of defects and fracture resistance of vessel materials." Strength of Materials 17, no. 11 (1985): 1481–89. http://dx.doi.org/10.1007/bf01529931.

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26

Erhard, Anton, and Frank Otremba. "Degradation of Material Properties Significant for Lifetime Extension of Nuclear Power Plants." Materials Testing 52, no. 1-2 (2010): 11–19. http://dx.doi.org/10.3139/120.110106.

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27

Kamaya, Masayuki. "Failure assessment curve for austenitic stainless steel pipes of nuclear power plants." Engineering Fracture Mechanics 238 (October 2020): 107283. http://dx.doi.org/10.1016/j.engfracmech.2020.107283.

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28

Raj, Baldev, D. K. Bhattacharya, and P. Rodriguez. "Development of in-service inspection techniques for nuclear power plants in India." International Journal of Pressure Vessels and Piping 56, no. 2 (1993): 183–211. http://dx.doi.org/10.1016/0308-0161(93)90093-9.

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29

Roos, E., K. H. Herter, and X. Schuler. "Lifetime management for mechanical systems, structures and components in nuclear power plants." International Journal of Pressure Vessels and Piping 83, no. 10 (2006): 756–66. http://dx.doi.org/10.1016/j.ijpvp.2006.07.008.

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30

Gorynin, I. V., Yu I. Zvezdin, B. T. Timofeev, V. M. Filatov, and V. A. Ignatov. "Fracture resistance of welded thick-walled high-pressure vessels in power plants. Report No. 2. Approach to evaluating static strength." Strength of Materials 17, no. 11 (1985): 1490–97. http://dx.doi.org/10.1007/bf01529932.

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31

HIGUCHI, Makoto, Takao NAKAMURA, and Yasuaki SUGIE. "Development of Environmental Fatigue Evaluation Method for Nuclear Power Plants in JSME Code." TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series A 76, no. 762 (2010): 171–81. http://dx.doi.org/10.1299/kikaia.76.171.

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32

Zheng, Yan-guo, and Hui-qiang Li. "Evaluation of protective quality of prestressed concrete containment buildings of nuclear power plants." Journal of Central South University of Technology 18, no. 1 (2011): 238–43. http://dx.doi.org/10.1007/s11771-011-0685-7.

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33

Moraga, N. O., D. L. Jacobson, and J. F. Morris. "Fracture-resistant ultralloys for space-power systems: Nuclear-thermionic-conversion implications of W,27Re." Engineering Fracture Mechanics 34, no. 3 (1989): 553–65. http://dx.doi.org/10.1016/0013-7944(89)90118-5.

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34

Savin, A., L. Udpa, R. Steigmann, R. Grimberg, and S. Udpa. "Nondestructive examination of fuel channels in PHWR nuclear power plants." International Journal of Materials and Product Technology 27, no. 3/4 (2006): 198. http://dx.doi.org/10.1504/ijmpt.2006.011270.

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35

Ozhigov, L. S., A. S. Mitrofanov, and V. N. Voevodin. "Corrosion of the Second Contour of Power-Generating Units of Nuclear Power Plants with WWÉR-1000 Reactors." Materials Science 52, no. 5 (2017): 654–60. http://dx.doi.org/10.1007/s11003-017-0005-y.

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36

Kanto, Y., K. Onizawa, H. Machida, Y. Isobe, and S. Yoshimura. "Recent Japanese research activities on probabilistic fracture mechanics for pressure vessel and piping of nuclear power plant." International Journal of Pressure Vessels and Piping 87, no. 1 (2010): 11–16. http://dx.doi.org/10.1016/j.ijpvp.2009.11.010.

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37

Gorynin, I. V., and B. T. Timofeev. "Aging of materials of the equipment of nuclear power plants after designed service life." Materials Science 42, no. 2 (2006): 155–69. http://dx.doi.org/10.1007/s11003-006-0068-7.

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38

Naik, Rajiv A., Daniel P. DeLuca, and Dilip M. Shah. "Critical Plane Fatigue Modeling and Characterization of Single Crystal Nickel Superalloys." Journal of Engineering for Gas Turbines and Power 126, no. 2 (2004): 391–400. http://dx.doi.org/10.1115/1.1690768.

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Single crystal nickel-base superalloys deform by shearing along 〈111〉 planes, sometimes referred to as “octahedral” slip planes. Under fatigue loading, cyclic stress produces alternating slip reversals on the critical slip systems which eventually results in fatigue crack initiation along the “critical” octahedral planes. A “critical plane” fatigue modeling approach was developed in the present study to analyze high cycle fatigue (HCF) failures in single crystal materials. This approach accounted for the effects of crystal orientation and the micromechanics of the deformation and slip mechanis

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39

Young, Lih-jier. "A fracture mechanics analysis of the PWR nuclear power plant reactor pressure vessel beltline weld." Journal of Nuclear Materials 288, no. 2-3 (2001): 197–201. http://dx.doi.org/10.1016/s0022-3115(00)00710-8.

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40

Park, June-soo, Ha-cheol Song, Ki-seok Yoon, Taek-sang Choi, and Jai-hak Park. "Structural integrity evaluation for interference-fit flywheels in reactor coolant pumps of nuclear power plants." Journal of Mechanical Science and Technology 19, no. 11 (2005): 1988–97. http://dx.doi.org/10.1007/bf02916491.

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41

Timofeev, B. T., and Zh L. Bazaras. "Cyclic strength of the equipment of nuclear power plants made of 22K steel." Materials Science 41, no. 5 (2005): 680–85. http://dx.doi.org/10.1007/s11003-006-0031-7.

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42

KAMAYA, Masayuki, Hideo MACHIDA, Masao ITATANI, and Kiminobu HOJO. "Z-factor equations for elastic-plastic fracture mechanics analysis prescribed in the JSME rules on fitness-for-service for nuclear power plants." Transactions of the JSME (in Japanese) 82, no. 841 (2016): 16–00263. http://dx.doi.org/10.1299/transjsme.16-00263.

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43

Vystavkin, I. A., V. M. Torop, and O. V. Biryukov. "Estimation of the state of the metal of pipelines of nuclear power plants after 100,000 hours of operation." Strength of Materials 30, no. 2 (1998): 145–51. http://dx.doi.org/10.1007/bf02811279.

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44

Neklyudov, I. M., V. M. Azhazha, L. S. Ozhigov, and A. S. Mitrofanov. "In-service damage to heat-exchange tubes and welded joints in steam generators of power-generating units of nuclear power plants with WWER-1000." Strength of Materials 40, no. 2 (2008): 241–45. http://dx.doi.org/10.1007/s11223-008-9007-1.

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45

Parshin, V. M., I. I. Sheinfel'd, M. G. Chigrinov, A. V. Larin, and A. M. Chigrinov. "Specialized Production of Tubular Semifinished Products for Nuclear Power Plants from Deactivated Scrap Metal." Metallurgist 48, no. 5/6 (2004): 229–32. http://dx.doi.org/10.1023/b:mell.0000042817.00611.28.

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46

Timofeev, B. T., V. A. Fedorova, and A. A. Buchatskii. "Intercrystalline Corrosion Cracking of Welded Joints of the Austenitic Pipelines of Nuclear Power Plants." Materials Science 40, no. 5 (2004): 676–83. http://dx.doi.org/10.1007/s11003-005-0099-5.

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47

Timofeev, B. T., and Zh L. Bazaras. "Low-Cycle Fatigue of the Equipment of Nuclear Power Plants Made of 15Kh2MFA Steel." Materials Science 41, no. 3 (2005): 410–17. http://dx.doi.org/10.1007/s11003-005-0179-6.

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48

Shibata, H. "Assumed Process of Piping Failure in Nuclear Power Plants Under Destructive Earthquake Conditions." Journal of Pressure Vessel Technology 113, no. 2 (1991): 268–72. http://dx.doi.org/10.1115/1.2928754.

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This paper deals with an assumed process of piping failure in nuclear power plants which may cause a catastrophic accident during a destructive earthquake condition. The type of failure discussed is the so-called double-ended guillotine break, DEGB. As a safety problem, we are going to eliminate this type of failure by LBB, and we have assumed that this would then not occur by an earthquake. The author tries to clarify the possibility of failure during earthquakes. He reviews his related papers since 1973, and discusses zipping failure of snubbers and supporting devices. He shows a procedure t

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49

Kamaya, Masayuki, and Hideo Machida. "Reference stress method for evaluation of failure assessment curve of cracked pipes in nuclear power plants." International Journal of Pressure Vessels and Piping 87, no. 1 (2010): 66–73. http://dx.doi.org/10.1016/j.ijpvp.2009.11.002.

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50

Huang, Yin-Nan, and Andrew S. Whittaker. "Vulnerability Assessment of Conventional and Base-Isolated Nuclear Power Plants to Blast Loadings." International Journal of Protective Structures 4, no. 4 (2013): 545–63. http://dx.doi.org/10.1260/2041-4196.4.4.545.

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