[40]
[41],[42]. After the contract researches sponsored by JAERI were completed in 2000, the research activities have been still continued voluntarily by FFM Sub-Committee in JWES
Google Scholar
[1]
Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, NUREG-1061(1984)
Google Scholar
[2]
U.S.NRC Regulatory Guide 1.154 Format and Content of Plant Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactor (1987)
Google Scholar
[3]
Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-S-2002 (2002)
Google Scholar
[4]
N.Soneda: Verification of Probabilistic Fracture Mechanics Analysis Code by PTS Benchmarking Study, CRIEPI-Report T93053 (1994)
Google Scholar
[5]
A basic Study on Material Degradation for Structural Components of Light Water Reactors (Part 1), JWES-AE-8901 (1989) (in Japanese)
Google Scholar
[6]
A basic Study on Material Degradation for Structural Components of Light Water Reactors (Part 2), JWES-AE-9003 (1990) (in Japanese)
Google Scholar
[7]
A basic Study on Material Degradation for Structural Components of Light Water Reactors (Part 3), JWES-AE-9103 (1991) (in Japanese)
Google Scholar
[8]
Study on Application of Probabilistic Fracture Mechanics Methodology(1), Report of RC 111 Committee, JSME (1993) (in Japanese)
Google Scholar
[9]
Study on Application of Probabilistic Fracture Mechanics Methodology(2), Report of RC 111 Committee, JSME (1994) (in Japanese) [10] Study on Application of Probabilistic Fracture Mechanics Methodology(3), Report of RC 111 Committee JSME (1995) (in Japanese)
Google Scholar
[11]
Study on Application of Probabilistic Fracture Mechanics on Reliability Evaluation of Nuclear Components (1), Report of PFM Sub-Committee, JWES-AE-9804 (1997) (in Japanese)
Google Scholar
[12]
Study on Application of Probabilistic Fracture Mechanics on Reliability Evaluation of Nuclear Components (2), Report of PFM Sub-Committee, JWES-AE-9904 (1998) (in Japanese)
Google Scholar
[13]
Study on Application of Probabilistic Fracture Mechanics on Reliability Evaluation of Nuclear Components (3), Report of PFM Sub-Committee, JWES-AE-0002 (1999) (in Japanese)
Google Scholar
[14]
Study on Application of Probabilistic Fracture Mechanics on Reliability Evaluation of Nuclear Components (4), Report of PFM Sub-Committee, JWES-AE-0002 (2000) (in Japanese)
Google Scholar
[15]
Study on Application of Probabilistic Fracture Mechanics on Reliability Evaluation of Nuclear Components (5), Report of PFM Sub-Committee, JWES-AE-0102 (2001) (in Japanese)
Google Scholar
[16]
K. Shibata, D.Kato and Y.Li: Introduction of Effect of Annealing into Probabilistic Fracture Mechanics Code and results of Benchmark Analyses, Presented in ASME PVP-Vol.400 (2000), Seattle, USA
Google Scholar
[17]
K.Shibata, K..Onizawa,Y.Li and and D.Kato: Development of Probabilistic Fracture Mechanics Code PASCAL and User's Manual, JAERI-Data/Code 2001-011, (2001)
Google Scholar
[18]
K.Shibata, D.Kato and Y.Li: Development of PFM Code for Evaluating Reliability of Pressure Components Subject to Transient Loading, Nucl. Eng. Des., 208 (2001), p.1
DOI: 10.1016/s0029-5493(01)00361-2
Google Scholar
[19]
Y.Li, D.Kato, K.Shibata and K.Onizawa: Development of PFM Code with a Function of Ductile Crack Extension Analysis, Trans JSME, Ser.A, Vol.69,No.678 (2003,p.463
Google Scholar
[20]
K.Shibata, K..Onizawa,Y.Li and D.Kato: Importance of Fracture Criterion and Material Characterization in Probabilistic Fracture Mechanics Analysis of an RPV under a Pressurized Thermal Shock, Int. Jour. Pres. Ves. Piping, 81 (2004), p.749
DOI: 10.1016/j.ijpvp.2004.05.003
Google Scholar
[21]
K.Onizawa, K.,Shibata, D.Kato and Y.Li: Probabilistic Fracture Mechanics Analysis of Reactor Pressure Vessel under PTS Transients: JAME Int. Jour., Series A, Vol.47, No.3 (2004), p.486
DOI: 10.1299/jsmea.47.486
Google Scholar
[22]
ASME Boiler and Pressure Vessel Code Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components (2001)
DOI: 10.1115/1.859872.ch29
Google Scholar
[23]
Code for Nuclear Power Generation Facilities-Rules on Fitness for Service for Nuclear Power Plants, JSME SNA1-2002 (2002)
Google Scholar
[24]
G.Yagawa, S.Yoshimura, N.Handa, et. al.:Japanese Round Robin Analysis for Probabilistic fracture Mechanics, Trans. SMiRT 11, Vol.G, (1991)
Google Scholar
[25]
S.Yoshimura, G.Yagawa, T.Iida, et.al.: Probabilistic Farcture Mechanics Analysis of Aged Structures and Materials, Jour. of Atomic Energy Society of Japan, Vol.34, No. 12 (1992)
Google Scholar
[26]
G.Yagawa, S.Yoshimura, N.Handa, et.al: Study on Life Extension of Aged RPV Material Based on Probabilistic Fracture Mechanics: Japanese Round Robin, Jour.of Press. Ves. Tech, Vol.111 (1995)
DOI: 10.1115/1.2842095
Google Scholar
[27]
N.Soneda, M.Hirano, Y.Ohtani, et. al.:Japanese PTS Benchmarking Study using Probabilistic Fracture Mechanics, Proc. of ICONE-3, Vol.1 (1995), Kyoto, Japan
Google Scholar
[28]
Y.Kanto, S.Yoshimura, H.Ueda, et. Al: Application of Probabilistic Fracture Mechanics to Pressure Vessels and Piping of Nuclear Power Plants: JSME Activity, Proc. of ICON-3, Vol.1 (1995), Kyoto, Japan [29] M.Hirano, N.Watanabe, et.al.: Comparison between VISA-II and OCA-P for Probabilistic Fracture Mechanics Analysis Focusing on analysis method, Trans. 13th SMiRT (1995),Port Alegre, Brazil
DOI: 10.1016/j.ijpvp.2013.10.010
Google Scholar
[30]
G.Yagawa, Y.Kanto and S.Yoshimura: Probabilistic Fracture Mechanics of Nuclear Structural Components: Consideration of Transition from Embedded Crack to Surface Crack, Nucl. Eng. Des., Vol.191 (1999), p.263
DOI: 10.1016/s0029-5493(99)00148-x
Google Scholar
[31]
Y.Kanto, G.Yagawa: Probabilistic Fracture Mechanics Analysis of Nuclear Reactor Pressure Vessels: Effects of Transition from Embedded Crack to Surface Crack, Trans 15 th SMiRT, Vol.X, (1999), Seoul, Korea
DOI: 10.1016/s0029-5493(99)00148-x
Google Scholar
[32]
Y.Kanto, D.Yagawa: Analysis of Embedded Cracks in RPV, Trans 16th SMiRT Paper # 1608 (2001),Washington DC, USA
Google Scholar
[33]
H.Machida, M.Arakawa and Y.Kamishima: Probabilistic Fracture Mechanics Analysis for Pipe Considering Dispersion of Seismic Loading, Nucl. Eng. Des., Vol. 212 (2002), p.1
DOI: 10.1016/s0029-5493(01)00469-1
Google Scholar
[34]
H.Machida and S.Yoshimura: Probabilistic Fracture Mechanics Analysis of Nuclear Piping Considering Variation of seismic Loading, Int. Jour. Pressure Vessel and Piping, Vol. 79 (2002), p.193
DOI: 10.1016/s0308-0161(02)00011-x
Google Scholar
[35]
H.Machida and S.Yoshimura: Probabilistic Fracture Mechanics Analysis of Nuclear Piping Considering an Embedded Crack, ASME PVP, Vol.443-2 (2002),Vancouver, Canada
DOI: 10.1115/pvp2002-1345
Google Scholar
[36]
N.Maeda, S.Nakagawa, G..Yagawa and S.Yoshimura: Calculation of Risk Informed Inspection and Cost Effective Maintenance Using Probabilistic Fracture Mechanics, Trans. 15 th SMiRT, Vol.V (1999), Seoul, Korea
Google Scholar
[37]
H.Machida, N.Yamashita, S.Yoshimura and G.Yagawa: Reliability Assessment of Piping in a Nuclear Power Plant Considering Flaw Detection Probability, ASME/JSME PVP-Vol.480 (2004), San Diego, USA
DOI: 10.1115/pvp2004-2698
Google Scholar
[38]
Y.Isobe, M.Sagisaka, S.Yoshimura and G.Yagawa: Risk-benefit analysis of SG tube maintenance based on probabilistic fracture mechanics, Nucl. Eng. Des., 207 (2001), p.287
DOI: 10.1016/s0029-5493(01)00338-7
Google Scholar
[39]
M.Sagisaka, Y.Isobe, S.Yoshimura and G.Yagawa: Quantitative Evaluation of Maintenance Strategies for Steam Generator Tubes Using Probabilistic Fracture Mechanics, Jour. of Atomic Energy Society of Japan, Vol.42, No.12 (2000), P.1325
DOI: 10.3327/jaesj.42.1325
Google Scholar
[40]
G.Yagawa, Y.Kanto, S.Yoshimura, H.Machida and K.Shibata: Probabilistic Fracture Mechanics Analysis of Nuclear Structural Components: A Review of Recent Japanese Activities, Nucl. Eng. Des., Vol.207 (2001), p.269
DOI: 10.1016/s0029-5493(01)00337-5
Google Scholar
[41]
G.Yagawa, Y.Kanto, S.Yoshimura and K. Shibata: Recent Research Activity of Probabilistic Fracture Mechanics for Nuclear Structural Components in Japan, Trans. 16 th SMiRT, Papeer♯ 1610 (2001), Washington DC, USA
DOI: 10.1016/s0029-5493(01)00337-5
Google Scholar
[42]
S.Yoshimura, G.Yagawa, Y.Kanto and K.Shibata: Probabilistic Fracture Mechanics Based on Assessments for Aged Nuclear Structural Components, Trans. 16th SMiRT, Papeer♯2076 (2001), Washington DC, USA
Google Scholar
[43]
Study on Application of Probabilistic Fracture Mechanics on Reliability Evaluation of Nuclear Components, Report of PFM Sub-Committee, JWES-AE-0204 (2002) (in Japanese)
Google Scholar
[44]
Study on Application of Probabilistic Fracture Mechanics on Reliability Evaluation of Nuclear Components, Report of PFM Sub-Committee, JWES-AE-0304 (2003) (in Japanese) [45] Study on Application of Probabilistic Fracture Mechanics on Reliability Evaluation of Nuclear Components, Report of PFM Sub-Committee, JWES-AE-0301 (2004) (in Japanese)
DOI: 10.1016/0143-8174(85)90026-5
Google Scholar
[46]
I.Milne, R.A. Ainsworth, A.R. Dowling, et.al.: Assessment of the Integrity of Structures Containing Defects, CEGB Report R/H/R6-Rev.3 (1986)
Google Scholar
[47]
M.Tomimatsu, S.Asada and H.Namatame: Evaluation of RPV Steel Surveillance Program in Japanese PWR: Radiation Embrittlement, Prediction, Int. Nat. Symposium on Reactor Dosimetory (1996), Plague
Google Scholar
[48]
E.D. Eason, J.E. Wright and E.E. Nelson: Multivariable Modeling of Pressure Vessel and Piping J-R Data, NUREG/CR-5729 (1991)
DOI: 10.2172/5822791
Google Scholar
[49]
USNRC Regulatory Guide 1.162: Format and Content of Report for Thermal Annealing of Reactor Pressure Vessels (1996)
Google Scholar
[50]
K. Balkey, F.J. Witt and B.A. Bishop: Documentation of Probabilistic Fracture Mechanics Codes Used for Reactor Pressure Vessels Subjected to Pressurized Thermal Shock Loading. EPRITR-105001 (1995) Table 1 Progress of PFM researches sponsored by JAERI* Period Executing Organization Components Main Activities Remarks References 1988 ~ 1990 JWES1) PFM WG Under LE Sub-Committee RPV ・Survey of existing PFM code and numerical and fracture mechanics models for PFM analysis ・PFM Round Robin Analysis of RPV ・Research on numerical algorithm
DOI: 10.1115/pvp2017-65225
Google Scholar
[7]
1991 MRI2) PFM Committee RPV Piping ・Survey of input data ・Survey of analysis model 1992 ~ 1994 JSME3) RC111 committee RPV Piping ・Survey of input data ・Survey of analysis model and research on numerical algorithm ・Round Robin Analysis of RPV under operating load and PTS ・Round Robin Analysis of Piping Proposal of a standard guideline for PFM analysis [8],[9]
DOI: 10.1016/j.ijpvp.2018.10.007
Google Scholar
[10]
1996 ~ 2000 JWES PFM Sub-Committee RPV Piping SG ・Refinement of PFM methodology: Input seismic load, SIF database, Treatment of embedded crack ・Application to ISI code ・Application to RII, cost/benefit analysis in inspection strategy ・Utilization of PASCAL by round robin analyses ・Survey of application in other fields, need in structural integrity issues in LWR components Practical application to structural integrity issues of LWR components
Google Scholar
[15]
2001 ~ * JWES* PFM Sub-Committee Ditto Ditto Ditto
Google Scholar