ML022820849

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Handouts for 09/18/2002 Meeting Summary with the Nuclear Energy Institute (NEI) on the Environmental Assisted Fatigue Issued for a License Renewal Application
ML022820849
Person / Time
Site: PROJ0690
Issue date: 09/18/2002
From: Robinson M R
Duke Energy Corp
To:
Office of Nuclear Reactor Regulation, Nuclear Energy Institute
Kang P J, NRR/DRIP/RLEP, 415-2279
References
Download: ML022820849 (65)
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Text

Aging Management of Environmental Fatigue for Carbon/Low-Alloy Steels Michael R.Robinson Duke Energy September 18, 2002 NRC Headquarters Enclosure 3 091802.ppt 1

Objectives

"* Present industry basis for resolving environmental fatigue issue for carbon and low-alloy steel components during license renewal "* Reach agreement with NRC management on a process for NRC review of industry basis "* Discuss aging management of fatigue for carbon and low-alloy steel during license renewal "* Establish post-meeting actions 091802.ppt 2

Discussion Outline "* Background

"* Aging management for fatigue of carbon and low-alloy steel components during license renewal "* Supporting presentations

"* Re-evaluation of NUREG/CR-6674

"* Review of laboratory and component/structural fatigue data

  • Conclusions
  • Actions 091802 ppt 3 Background
  • Significant industry activities over the past decade to investigate the effects of reactor water environment on fatigue life of metal components.
  • Since mid-2000 the effort has been coordinated under the EPRI MRP Fatigue Issue Task Group (ITG): "* Focal point for industry technical efforts "* Technically represents NEI License Renewal Working Group "* Provides the technical interface with NRC staff "* Provides guidance to license renewal applicants

"* Provides results to ASME Code for consideration

  • MRP environmental fatigue program structured to address near-term industry needs and resolve long-term technical issues 091802 ppt 4 Issue Definition
  • NRC Staff and industry have yet to reach consensus on the need to explicitly account for the effects of reactor water environment as part of an overall fatigue management program during the license renewal period 091802 ppt 5 Background
  • Environmental fatigue found by NRC to be risk insignificant and no generic regulatory action required for:

"* Most high fatigue locations could be excluded with more detailed analyses "* Fatigue failure of piping is an insignificant contributor to core melt frequency 091802 ppt 6 Background

  • Environmental fatigue found by NRC to be risk insignificant and no generic regulatory action required for:
  • 60-year operating life (Thadani GS!-190 closeout memorandum to Travers, December 26, 1999) "* ALWR Sufficient conservatism in the fatigue analyses performed for the generic 60-year ALWR life to account for environmental effects.

"* License renewal period of existing plants (40-60 years) vNUREG/CR-6674 evaluated effect of environmental fatigue on overall risk through 60 years of operation v Concluded that environmental fatigue not a risk-significant issue for 20 years of additional operation

  • Predicted increase in frequency of pipe leakage in 40-60 year period 091802.ppt 7

Current Status

  • Because of increased leakage potential, the closeout of GSI-190 requires license renewal applicants to address environmental fatigue in aging management programs.
  • Significant resources expended by industry and NRC to address this issue
  • NRC Staff and industry have yet to reach consensus on the need to explicitly account for the effects of reactor water environment as part of an overall fatigue management program during the license renewal period 091802.ppt 8

MRP Program Activities

"* Near-term guidance to address environmental fatigue in a license renewal application (MRP-47 published October 2001) "* Previous applicants utilized different approaches

"* Consistent application of previous approaches desired to simplify process "* RAIs received June 26, 2002; responses under development

"* MRP-47 assumes consideration of environmental fatigue is necessary (until results of long-term efforts are known) "* Long-term activities to address identified technical issues "* Assess relevance of previous work on environmental fatigue to current plant operating conditions

"* Determine scope of aging management program necessary to address environmental fatigue in license renewal.

"* Results of long-term activities may necessitate change in near term guidance document (MRP-47).091802 ppt 9 Ongoing Long-Term MRP Activities

"* Review assumptions and inputs to the original risk study for conservatism and applicability

"* Review basis for determining significance of predicted component leakage "* Re-evaluate components for changes in leakage probabilities using revised assumptions and other inputs

  • Review and evaluate available environmental fatigue data
  • Reconcile available laboratory data with structural/component test results and plant operating experience (MRP-49 published in December 2001)091802 ppt 10 Current Aging Management Approach
  • Plants rely on existing programs to effectively manage fatigue "* Compliance with current fatigue licensing basis through: "+ Cycle counting and comparison to design limits "* Fatigue monitoring to count and categorize thermal cycles "* Re-analysis, if necessary, to account for actual cycles and transient severity "* Structural integrity of fatigue-sensitive locations

"* Formal inservice examination requirements provided in each plant's inservice inspection programs "* Augmented inspection/evaluation, if necessary, based on plant operating experience and/or regulatory enforcement actions "* Inspection scope and frequency expanded if flaws detected "* Risk-informed considerations now developed

  • The industry believes this approach is adequate for license renewal period 091802 ppt 11 Basis for Continuing Current Aging Management Approach During License Renewal e Results from MRP long-term activities indicate no need for any formal consideration of environmental fatigue for carbon and low-alloy steel components
  • Results of NUREG/CR-6674 risk study re-analysis (MRP-74) " Fatigue failure is an even less significant contributor to increases in core damage frequencies

(<10-10) and is well below the NRC threshold (10-6) for being risk significant

"* Several orders of magnitude reduction in crack initiation and leakage probabilities

"+ Predicted 60-year leakage probabilities are not significant and are below previously NRC accepted leakage probabilities at 40 years in NUREG/CR-6674 v Maximum 40-year leakage probability from NUREG/CR-6674

,/ 0.41 Maximum 60-year leakage probability from MRP re-evaluation

,. 0.0014 091802.ppt 12 Basis for Continuing Current Aging Management Approach During License Renewal

  • Results from MRP long-term activities indicate no need for any formal consideration of environmental fatigue for carbon and low-alloy steel components

"* MRP evaluation of available laboratory, component, and structural data generated under simulated reactor water environmental conditions showed behavior consistent with margins in ASME Code fatigue design curve "* Flow rate identified as critical variable, but not simulated in typical laboratory environmental fatigue tests "* Component/structural tests are more representative of actual plant operating conditions (oxygen content, flow rate, size effects, surface finish) "* Existing fatigue design process is sufficiently conservative to account for a potential environmental effect "* ASME Code methods "* Design transient definition, number of cycles and severity 091802 ppt 13 Conclusions

"* MRP evaluation concludes consideration of environmental fatigue effects for carbon and low alloy steel components, as stipulated in GSI-190 closeout memorandum, is not warranted

"* All carbon/low-alloy steel fatigue locations can continue to rely on existing plant programs to track component fatigue usage through the license renewal period and remain in compliance with all NRC regulatory requirements

"* MRP evaluation of austenitic stainless steel locations is continuing 091802.ppt 14 RE-EVALUATION OF RESULTS FROM NUREGICR-6674 "FATIGUE ANALYSIS OF COMPONENTS FOR 60-YEAR PLANT LIFE" FOR FERRITIC STEEL COMPONENTS Presentation to NRC Art Deardorff Structural Integrity Associates September 18, 2002 Enclosure 4 091802.ppt 16 OUTLINE "* Review NUREG/CR-6260 fatigue analysis

  • Application of NUREG/CR-5999 "Interim Fatigue Curves to Selected Nuclear Power Plant Components" published March 1995 "* Review NUREG/CR-6674 probabilistic evaluation
  • Fatigue Analysis of Components for 60-Year Plant Life published June 2000 "* Present results of re-evaluation for carbon and low alloy components

"* Provide conclusions that core damage frequencies and leakage probabilities much less than previously documented 091802.ppt 17 BACKGROUND-NUREG/CR-6260

"* NUREG/CR-5999 (ANL 1993) proposed modified fatigue curves to consider environmental effects for carbon, low alloy and austenitic stainless steels "* Idaho National Engineering Laboratories (INEL) contracted by NRC "* Obtained stress analyses for six representative components for seven older/newer plant types "* Performed analysis for some piping where design was per ANSI B31.1 "* Assessed effect of modified curves for 40 years "* Where CUF high removed conservatism/considered actual cycles "* Projected results to 60 years (multiplier of 1.5 x 40-year CUF) "* Could not explicitly show that for all components CUF < 1.0

  • Stated that detailed transient monitoring or refined analysis could be used to demonstrate CUF < 1.0 091802 ppt 18 BACKGROUNDm-NUREG/CR-6260 Results for ferritic components NUREG/CR-5999 CUF VENDOR/VINTAGE LOCATION 40 YR DESIGN CUF CONSERVATISMS DESIGN REMOVED AND/OR E 60OYR EXPECTED CYCLES CE NEW RPV LOWER HEAD/SHELL 0.007 0.014 0.021 CE NEW INLET NOZZLE 0.182 0.475 0.712 CE NEW OUTLET NOZZLE 0.377 0.835 0.472 0.708 CE NEW CHARGING NOZZLE 0.05 0.104 0.156 CE NEW SAFETY INJECTION 0.898 2.101 0.457 0.686 CE OLD RPV LOWER HEAD/SHELL 0.008 0.013 0.02 CE OLD INLET NOZZLE 0.073 0.172 0.258 CE OLD OUTLET NOZZLE 0.284 0.554 0.831 B&W RPV LOWER HEAD/SKIRT 0.12 0.223 0.335 B&W OUTLET NOZZLE 0.9 2.148 0.469 0.704 B&W HOT LEG SURGE NOZZLE 0.592 1.092 0.47 0.705 B&W CORE FLOOD NOZZLE 0.345 0.632 0.948 W NEW RPV LOWERHEAD 0.012 0.018 0.027 W NEW INLET NOZZLE 0.11 0.29 0.435 091802 ppt 19 BACKGROUND-NUREGICR-6260 Results for ferritic components (continued)

NUREG/CR-5999 CUF VENDOR/VINTAGE LOCATION 40 YR DESIGN CUF CONSERVATISMS DESIGN REMOVED AND/OR E 60OYR EXPECTED CYCLES W NEW OUTLET NOZZLE 0.398 0.658 0.987 W OLD RPV AT SUPPORT 0.29 0.891 1.33 W OLD INLET NOZZLE 0.28 0.496 -0.744 W OLD OUTLET NOZZLE 0.431 1.161 0.347 0.52 GE NEW RPV LOWER HEAD 0.2 11.702 0.628 0.942 GE NEW FW NOZZLE SAFE-END 0.301 1.73 1.881 2.822 GE NEW CORE SPRAY SAFE-END 0.05 0.675 0.436 0.654 GE NEW RHR LINE PIPE 0.407 11.26 16.89 GE NEW FEEDWATER ELBOW 0.435 3.746 3.688 5.532 GE OLD RPV LOWER HEAD 0.032 2.063 0.079 0.119 GE OLD FW NOZZZLE BORE 0.7 9.859 3.168 4.752 GE OLD CORE SPRAY NOZZLE 0.023 0.441 0.52 0.78 GE OLD FW LINE RCIC TEE 0.427 5.016 6.98 10.47 091802.ppt 20 BACKGROUND=-

NUREG/CR-6674

"* Pacific Northwest National Laboratories (PNNL) assessed the effects of water environment for 60 year life: leakage + core damage frequency

"* INEL results for stresses/environmental conditions

"* Fatigue curves for carbon steel/LAS taken from NUREG/CR-6335 (1995) "* Used an enhanced version of the pc-PRAISE code to evaluate probabilities of CUF exceeding 1.0 and through-wall cracking "* Through-wall stresses and component geometry (diameter/thickness) based on reasonable assumptions; NUREG/CR-6260 had only surface stress amplitudes

"* Key conclusions

"* Some components had through-wall crack probabilities

= 0.05/year

"* Probabilities of throughwall cracking in water environment approached 1.0 for some components for both 40 and 60 years "* Core damage frequencies

< 10-6 per year; much less in most cases 091802.ppt 21 BACKGROUNDm-NUREG/CR-6674 Results for ferritic components 091802.ppt 22 NUREG/CR-5999 NUREG/CR-6674 PROBABILITIES CUF 40-YEAR AIR 40-YEAR WATER 60-YEAR WATER VENDOR/VINTAGE LOCATION 40 YR DESIGN CUF 60 YR PTW PTW PTW EXTRAPOLATION CRACK CRACK CRACK CE NEW RPV LOWER HEAD/SHELL 0007 0.021 8.40E-23 1.65E-24 6.71E-15 1.13E-14 1.44E-12 1.91E-14 CE NEW INLET NOZZLE 0182 0.712 1.75E-07 6 94E-14 5 90E-05 2 03E-11 9.01E-04 2 05E-10 CE NEW OUTLET NOZZLE 0 377 0.708 1.00E-07 6 75E-14 1.74E-03 9 65E-10 2.90E-02 6.93E-09 CE NEW CHARGING NOZZLE 005 0.156 6 48E-11 8 32E-17 2.61E-06 2.77E-12 5.50E-05 4.05E-11 CE NEW SAFETY INJECTION 0898 0686 1.22E-06 8.30E-13 1.00E-06 1.88E-12 1.90E-05 7.50E-12 CE OLD RPV LOWER HEAD/SHELL 0.008 0.02 4.85E-24 9.56E-26 6.36E-16 1.08E-17 1.85E-13 1.86E-14 CE OLD INLET NOZZLE 0073 0258 7.99E-11 3 40E-17 4.11E-07 1.58E-13 1.33E-05 3.59E-12 CE OLD OUTLET NOZZLE 0284 0831 6 72E-04 1.91 E-10 7.05E-02 2 42E-08 3.53E-01 6.13E-08 B&W RPV LOWER HEAD/SKIRT 012 0.335 4.07E-09 6 24E-11 7.85E-06 1.04E-07 7.52E-04 1.36E-06 B&W OUTLET NOZZLE 0.9 0.704 2.92E-03 6.72E-10 1.83E-01 5 25E-08 4.55E-01 9.03E-08 B&W HOT LEG SURGE NOZZLE 0592 0.705 ESTIMATED CUF -STRESSES NOT AVAILABLE B&W CORE FLOOD NOZZLE 0345 0948 NOT ANALYZED SINCE STAINLESS STEEL SAFE-END LOCATION MORE B&W__COREFLOODNOZZLE_

0_345_0_948CONTROLLING W NEW RPV LOWERHEAD 0.012 0027 1.04E-19 1.80E-21 7.52E-13 1.24E-14 964E-11 1.50E-12 W NEW INLET NOZZLE 011 0435 7.03E-10 3.OOE-16 9.17E-07 3.51E-13 2 84E-05 7.64E-12 BACKGROUND=-

NUREG/CR-6674 Results for ferritic components (continued)

NUREGICR-5999 NUREGICR-6674 PROBABILITIES CUF 40-YEAR AIR 40-YEAR WATER 60-YEAR WATER VENDORIVINTAGE LOCATION 40 YR DESIGN CUF 60YR PTW PTW PTW EXTRAPOLATION CRACK CRACK CRACK W NEW OUTLET NOZZLE 0.398 0987 3 OOE-04 6.76E-1 1 3 65E-01 8.57E-08 7.42E-01 1.22E-07 W OLD RPV AT SUPPORT 029 1.33 1.06E-09 1.38E-1 1 7.20E-07 8.44E-09 1.11 E-05 9 15E-08 W OLD INLET NOZZLE 0 28 0.744 4.48E-04 1.28E-10 4 38E-03 2.03E-09 5 04E-02 1.07E-08 W OLD OUTLET NOZZLE 0.431 0.52 7.77E-03 1.89E-09 9 33E-03 4.21 E-09 9 60E-02 2.04E-08 GE NEW RPV LOWER HEAD 02 0942 3 59E-18 6.03E-20 7.88E-12 1.25E-13 6 82E-10 8 26E-12 GE NEW FW NOZZLE SAFE-END 0.301 2.822 2.OOE-06 1.88E-14 1.31E-03 3.57E-11 1.47E-02 1.84E-10 GE NEW CORE SPRAY SAFE-END 005 0654 6 54E-12 3.67E-19 1.45E-07 7.09E-1 5 3 25E-06 1.09E-13 GE NEW RHR LINE PIPE 0.407 1689 2.08E-01 1.35E-11 4.1OE-01 2.54E-11 6.21E-01 2.03E-10 GE NEW FEEDWATER ELBOW 0.435 5532 1.OOE-05 2.25E-1 1 1.01 E-03 3.04E-09 1.46E-02 5 06E-09 GE OLD RPV LOWER HEAD 0.032 0.119 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 GE OLD FW NOZZZLE BORE 07 4.752 4.53E-05 8.69E-14 1.00E-05 3.75E-14 8 80E-04 1.46E-12 GE OLD CORE SPRAY NOZZLE 0.023 078 4.44E-14 1.72E-22 1.91E-08 6.41 E-17 8 84E-07 2.14E-15 GE OLD FW LINE RCIC TEE 0.427 10.47 4.30E-06 2 OOE-13 2.99E-03 1.04E+10 5 92E-02 8 30E-10 091802.ppt 23 BACKGROUND

-GSI 190 CLOSURE "* Generic Safety Issue 190 was closed following December 26, 1999 memo from Thadani to Travers "* Low core damage frequencies were predicted even with most recent fatigue test data "* However, since studies indicated an increase in leakage frequency, recommendation was that licensees address environmental effects in support of license renewal "* Actions by NRC were reasonable based on conservative results presented in NUREG/CR-6674 m High probability of leakage from some components 091802 ppt 24 RE-EVALUATION OF NUREG/CR-6674

"* Project initiated to re-evaluate fatigue initiation and leakage probabilities, and core damage frequencies presented in NUREG/CR-6674

"* Re-evaluation included "* Critical review of input to NUREG/CR-6674

"* NRC version of pc-PRAISE further modified by Engineering Mechanics Technology

  • Benchmarked against NUREG/CR-6674 results "* Revised initiation/leakage probability predictions

"* Initial effort concentrated on carbon and low-alloy steel components 091802 ppt 25 DIFFERENCES FROM NUREG/CR-6674

"* Standard deviation at high-cycle end of fatigue curve was reduced "* Available data suggest standard deviation-0.1 Salt at 106 cycles "* NUREG/CR-6674 assumed standard deviation of 0.325 Salt for LAS and 0.277 Salt for CS (from NUREG/CR-6335)

  • Represents physically impossible material behavior "* Significantly affects components with large number of low-stress cycles m e.g., RPV nozzles designed for daily load following 091802.ppt 26 RE-EVALUATION OF NUREG/CR-6674
  • Low-alloy steel fatigue curve from NUREG/CR-6674 High Oxygen, 550 0 F, 0.01 %/sec Strain Rate, High Sulfur 1000.00 U-1 .100.00 10.00--- 0.0001 -U- 0.001 --0.01 --0.1 -"')--0.5 0.9 --- -0.99 -0.999 -0.9999 1000000 091802.ppt 27 10 100 1000 10000 100000 Number of Cycles RE-EVALUATION OF NUREG/CR-6674
  • Low-alloy steel fatigue curve from NUREG/CR-6674 with modified endurance limit variance High Oxygen, 550 0 F, 0.01 %/sec Strain Rate, High Sulfur 1000.00 1 . .= 100.00 (0 10.00-0.0001 --- 0.001 -- -0.01 ---0.1 -)K(- 0.5 "-4-0.9 ----0.99 -0.999 -0.9999 100 1000 10000 100000 Number of Cycles 091802.ppt 28 10 1000000 DIFFERENCES FROM NUREG/CR-6674

"* Latest environmental test data were considered Original work based on NUREG/CR-5999 (1993) and NUREG/CR-6335 (1995) "* NUREG/CR-6583 published March 1998 "* NUREG/CR-6717 published May 2001 "* New curves are more penalizing at low-cycle end and slightly less penalizing at high-cycle end "* All carbon and low-alloy steel components were re evaluated using modified endurance limit and updated fatigue curves 091802.ppt 29 RE-EVALUATION OF NUREG/CR-6674

  • Low-alloy steel fatigue curve with modified variance/latest data fit High Oxygen, 550 0 F, 0.01 %/sec Strain Rate, High Sulfur 1000.00 = 100.00 10.00 10 100 1000 10000 100000 Number of Cycles--- 0.0001 --0.001 -0.01 --E- 0.1 -)- 0.5 -0.9 --0.99 0.999 1- 0.9999 1000000 091802 ppt 30 RE-EVALUATION OF NUREG/CR-6774
  • Effect of updated fatigue data and endurance limit modification on probability of initiation at 60 years 1.E+O0"4 E .M ow a)r' >' E QD z 1.E-01 1.E-02 1.E-03 1 .E-04 1 .E-05 1.E-05 1.E-04 1 .E-03 1 .E-02 Cumulative Probability of Initiation (NUREG/CR-6674 Prediction) 091802 ppt 31'K N N N' N' N' N K' N V K N K' N 1.E-01 1.E+O0 RE-EVALUATION OF NUREG/CR-6674
  • Effect of updated fatigue data and endurance limit modification on probability of leakage at 60 years 1.E+O0 E CUr S-J~ .0 a.z 1.E-01 1.E-02 1.E-03 1 .E-04 1 .E-05 1.E-06 1.E-06 1 .E-05 1 .E-04 1.E-03 Cumulative Probability of Leakage (NUREG/CR-6674 Prediction) 091802.ppt 32K 'K 'K' i'K A' 'K ', I "K ' ' 'K "'K' -' 'K 'K 'K" ""K' '"K 'K "'K 'K "' 'K' K * '.' '4 ' 'K'" " " K ' "" 'K 'K '4'K."K.

""K ' "K 'K 1' 'K ' ' ' 'K <<'K \'K 'K 4' 4 ' ' ' 'K K" 'K ' ' ' 'K 'K 'K .'. "'K 'K ""'K" 'K K'" ' ' 'K 'K" 'K 'K 'K "K V 'K " ' 'K' 'K ' 'K 'KI 'K 'K 'K 'K",' K 'K'K'K 'K" ' I ' ' " 'K ' ' ,' ' 'K , 'K 'K'. 'K' 'K 'K "' 'K' 'K "K 'K' 'K' 'K'K', 'K 'K'K ' 'K 'K, 'K K 'K ' 'K ' "'K" 'K','K'K 'K 'K 'K 'K'K 'K '. ,'K 'K 'K" 'K 'K"' ' 'K, "K K 'K ' 'K .1 ' ,'KK ' 'K 'K 'K 'K' 'K 'K ' 'K' 'K , 'K' 'K 'K' 'K 'K' 'K 'K 'K , 'K "K 'K ' 'K 'K 'K "'K 'K 'K 'K 'K ""' 'K"'"' 'K"" 'K 'K 'K 'K 'K 'K 'K 'KK'1 .E-02 1 .E-01 1.E+O0 Further Refinements From NUREG/CR-6674

  • Locations with predicted leakage probabilities>

10-3 were further evaluated

  • Actual location operating temperatures do not approach 590OF for these components
  • For locations where detailed stress reports were available, actual geometry, stresses, strain rates and reduced cycles were used 091802 ppt 33 RE-EVALUATION OF NUREG/CR-6674
  • Detailed analysis of RPV outlet nozzles conducted to incorporate available stress report and cycle information Older CE Plant B&W Plant Probability Probability Case NUREG/CR-6717

+ All NUREGICR-6674 Endurance Limit Modifications Modified Initiation 40 Years 0.591 1.1 xl10-4 <10-6 60 Years 0.846 2.2 x 10-4 <10-6 Leakage 40 Years 0.071 <10.6 <106 60 Years 0.353 2.0 x 10.5 <10-6 Probability Probability Case NUREGICR-6717

+ All NUREG/CR-6674 Endurance Limit Modifications Modified Initiation 40 Years 0.774 6 x 10.6 <106 60 Years 0.899 1.4 x 10s <1 06 Leakage 40 Years 0.183 2 x 106 <10.6 60 Years 0.544 3 x 10.6 <10.6 091802.ppt 34 RE-EVALUATION OF NUREG/CR-6674

+ Consideration Probability Case NUREG/CR-6674 Endurance Limit of Actual Modified Temperatures Initiation 40 Years 0.104 0.0139 0.0004 60 Years 0.253 0.0533 0.0024 Leakage 40 Years 0.0013 0.0001 1 x 10.6 60 Years 0.0147 0.0017 3.5 x 105 Probability NUREG/CR-6717

+ Consideration Probability Case NUREG/CR-6674 Endurance Limit of Actual Modified Temperatures Initiation 40 Years 0.159 0.140 0.0032 60 Years 0.365 0.434 0.0192 Leakage 40 Years 0.0010 0.0003 2 x 10.6 60 Years 0.0146 0.0107 1.8 x 10-4 091802 ppt 35 MEMMESM RE-EVALUATION OF NUREG/CR-6674

+ Consideration of Actual Probability Case NUREGICR-6674 Endurance Limit Temperatures, OBE Modified Cycles and Strain Rates Initiation 40 Years 0.376 0.791 0.025 60 Years 0.782 0.981 0.108 Leakage 40 Years 0.0030 0.0169 6 x 105 60 Years 0.0592 0.2190 0.00139 091802.ppt 36 RE-EVALUATION OF NUREG/CR-6674

  • Final 60-year initiation probabilities 1 .E+O0 cn C. 0 cc (m 0 I IL 0 1.E-01 CL (U 1 .E-02 1.E-03 1.E-04 1.E-05 1.E-06 1.E-06 1 .E-05 1.E-04 1 .E-03 1 .E-02 Cumulative Probability of Initiation at 60 Years NUREG/CR-6674 091802 ppt 37' .... .. .."i 7;! : ; i'i > ~ i: ', S.... ..... ........" ...... ..... ... + " i ~ -7,.,. "

Vii I i '. 7, ! .."7 '-7 ... ! i ~ ! i '7' ' , A"i .!i ! ; ~ ii~ ! *. ... ii£!

i i i~~ii! ,i i 77 I S ,' i ," '7" 1.E-01 1.E+O0 RE-EVALUATION OF NUREG/CR-6674 9 Final 60-year leakage probabilities 1.E+00 1.E-01 1.E-02 1.E-03 1 .E-04 1 .E-05 1 .E-06 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 Cumulative Probability of Leakage at 60 Years NUREG/CR-6674 091802.ppt 38 ca 2... "C6 a, O a) I-tm C13 .x 0 2. .0. L.IL 0"Z, Ill w 1.E+00 RE-EVALUATION OF NUREG/CR-6674

.Comparison of revised analysis to NUREG/CR-6674 results 1.E+00 1.E-01 1.E-02 1.E-03 1.E-04 1.E-05 1.E-06 1 .E-07 1 .E-07 1.E-06 1.E-05 1.E-04 1.E-0ý3 -02 Cumulative Probability of Leakage at 40 Years NUREG/CR-6674 091802.ppt 39 a, 4.' .>_ C)uNUREG/CR-6674 A MRP Re-Analysis

-1:1 Line______

Probability-f leakage-i insignfican 1.E-01 1.E+00 RE-EVALUATION OF NUREG/CR-6674 e Core damage frequency re-evaluation 091802 ppt 40 PNNL Results Re-Evaluation Plant/Location Air Water Water CDF(40) CDF(40) CDF(60) CDF(40) CDF (60) B&W RPV OUTLET NOZZLE 6.72E-10 5.25E-08 9.03E-08 0 0 CE-NEW RPV OUTLET NOZZLE 6.75E-14 9.65E-10 6.93E-09 0 0 CE-NEW SAFETY INJECTION NOZZLE 8.30E-13 1.88E-12 7.50E-12 0 0 CE-OLD RPV OUTLET NOZZLE 1.91 E-10 2.42E-08 6.13E-08 0 0 GE-NEW FEEDWATER NOZZLE SAFE END 1.88E-14 3.37E-11 1.84E-10 2.12E-14 5.23E-13 GE-NEW RHR LINE STRAIGHT PIPE 1.35E-11 2.54E-11 2.03E-10 4.13E-14 2.26E-13 GE-NEW FEEDWATER LINE ELBOW 2.25E-1 1 3.04E-09 5.06E-09 2.70E-1 1 7.50E-1 1 GE-OLD RPV FEEDWATER NOZZLE BORE 8.69E-14 3.75E-14 1.46E-12 3.08E-16 2.65E-14 GE-OLD FEEDWATER LINE -RCIC TEE 2.OOE-1 3 1.04E-1 0 8.30E-1 0 8.77E-1 2 2.25E-1 1 W-NEW RPV OUTLET NOZZLE 6.76E-11 8.57E-08 1.22E-07 4.06E-13 2.71 E-12 W-OLD RPV INLET NOZZLE 1.28E-10 2.03E-09 1.07E-08 8.28E-14 5.78E-13 W-OLD RPV OUTLET NOZZLE 1.89E-09 4.21 E-09 2.04E-08 2.70E-1 3 0 RE-EVALUATION OF NUREG/CR-6674 9 Revised core damage frequency 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 CDF at 60 Years -NUREGICR-6674 1.E-08 1 .E-07 1.E-06 1.E-07 S1.E-08 LL S1.E-09 0 E 0 1.E-10 w .I 1.E-11 "o 1.E-12 LL (U a) cc* Versus Water Environment NN N x Versus Air N 1:1 LineN~:~ -----------------------------------'N 4 N' NN NN X , X'",NN N 4' 4x N ' N N" X -~- ---------' -- ---4 > x' T"4' X 0N>1.E-13 1.E-14 1.E-14 091802.ppt 41 1.E-06 CONCLUSIONS

"* Probabilities of initiation and leakage significantly less than reported in NUREG/CR-6674

"* More realistic fatigue curve endurance limit variance "* Use of latest environmental fatigue test data fits "* Use of actual temperatures

"* Examination of detailed stress analysis showed that probabilities could be further reduced "* 60-year leakage probability considering environmental effects for ferritic components not significant

"* Less than 40-year values from NUREG/CR-6674 accepted by NRC "* Maximum 60-year leakage probability of 0.0014 "* Maximum core damage frequency

< 1 x 1O-°1 0 year 091802.ppt 42 Review of Laboratory, Component and Structural Environmental Fatigue Data R. E. Nickell Applied Science & Technology, Poway, CA September 18, 2002 NRC Headquarters Enclosure 5 091802 ppt 44 Objectives

  • Demonstrate that data are consistent with the revised probabilities of through-wall cracking (and leakage) for component locations evaluated in NUREG/CR-6674

"* Show that laboratory test data obtained under simulated reactor water environmental conditions for carbon and low alloy steels are within margins in ASME Code fatigue design curve "* Show that structural/component fatigue test data with one surface in contact with water environment exhibit behavior consistent with margins in ASME Code fatigue design curve "* Show that operating experience has not revealed significant increases in fatigue failures ascribed to reactor water environmental effects as a function of increasing length of service of nuclear power plant components 091802.ppt 45 Carbon/Low-Alloy Steel Laboratory Data (Environmental Shift, No Size/Roughness Effect)10.00 1.00 0.10 0.01 4 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 Cyclic Life 091802 ppt 46 X0 0. E cn 1.E+06 Evaluation of Laboratory Data "* EPRI published a comprehensive review of laboratory data and the relationship to component/structural fatigue tests in December 2001 (MRP-49) "* Performed critical review of laboratory testing/environments and reconciled them with operating experience

"* Available structural/component test data are consistent with the majority of the laboratory test data and are within margins in ASME Code fatigue design curve; size effects and surface finish effects reduce the fatigue life only slightly "* Flow rate (rather than the trickle flow for most laboratory testing) has pronounced beneficial effect in operating environments 091802 ppt 47 Evaluation of Laboratory Data 9 Findings for Carbon Steel Laboratory Data and Component-Scale Tests "* Laboratory simulated reactor water data are within the margin in ASME Code fatigue design curve for environmental effects "* PVRC component-scale carbon steel tests demonstrate additional shift for size effects and surface finish effects "* Component-scale flow rate tests with trickle flow compromise full ASME Code margin of 20 at low-cycle end of design curve "* Moderate flow rate carbon steel component-scale tests shift fatigue life to appropriate position to the right of the ASME Code design curve 091802.ppt 48 Carbon Steel Laboratory Air Data (ASME Code Background Document)1.E+07 1.E+06 CL 0. 4L Cl) II Co 1.E+05 1.E+04 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 N, Cycles 091802.ppt 49 1.E+07 Carbon Steel Simulated PWR Data (PVRC Data at 288 0 C) 10.00 *CS-Simulated PWR --- ASME CS Mean Air, 2880C "1-ASME Design, 2880C 0- 1.00 E un 0.10 0.01 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06 1.E+07 Cyclic Life 091802.ppt 50 Carbon Steel Simulated BWR Data (PVRC Data, All Conditions)

I i-A--G A 0A Acon& A A 0DA AO 00 1.E+05 Cyclic Life 091802.ppt 51 10.00 0 E 1n 1.00 0.10+ CS-Simulated BWR, 250C

  • CS-Simulated BWR, 1000C o CS-Simulated BWR, 1500C E3 CS-Simulated BWR, 200 0 C
  • CS-Simulated BWR, 250 0 C o CS-Simulated BWR, 288 0 C ---ASME CS Mean Air, 288 0 C --ASME Design, 2880C So0 Ok S 0 .%Ol O 0.01 00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+06 1.E+07 1.E+08 1.E+09)

PVRC Carbon Steel Component Tests (Environmental Shift + Size/Roughness Effect)0-A x F 2X f ,f , ,-t ...i -i+ -I-i i 1 .E+04 1 .E+05 1.E+06 Number of Cycles 091802 ppt 52 100 90 80 C 0 0 M E ~0 0 X N 0 0 A201 Initiation X Failure O A201 Initiation and Failure ---ASMEOCS Mean Air ASME Design 70 60 50 40 30 20 10 04 v E0 1 .E+03 i % i r--l N-1 w 084 i I ft, V -% -

Evaluation of Laboratory Data

  • Findings for Low-Alloy Steel Laboratory Data and Component-Scale Tests "* Laboratory simulated reactor water data are within the margin in ASME Code fatigue design curve for environmental effects "* PVRC component-scale low-alloy steel tests demonstrate additional shift for size effects and surface finish effects 091802 ppt 53 Low-Alloy Steel Laboratory PWR Data 10.00 0 1.00 i ,n u.IU n Al 1.E+04 Cyclic Life 091802.ppt 54* LAS-Simulated PWR ---ASME LAS Mean Air, 2880C ASME Design, 2880C _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ] _ _ _ _ I _ _ _ _* *S.. ----I 1 .E+01 1 .E+02 1.E+03 1.E+05 1.E+06 1 .E+07 Low-Alloy Steel Laboratory BWR Data*01 1 1 T T 3 13.4ý3" 330 1.E+03 1.E+04 1 .E+05 1 .E+02 r'1.E+06 r 1.E+07 1.E+08 Cyclic Life 091802.ppt 55 10.00 C 1.00 0.10* LAS-Simulated BWR, 500C w LAS-Simulated BWR, 100'C A LAS-Simulated BWR, 1500C o LAS-Simulated BWR, 2000C o LAS-Simulated BWR, 2500C o3 LAS-Simulated BWR, 2880C ---ASME LAS Mean Air, 2880C -ASME Design, 2880C 0.01 4. 1.E+1.E+09 1 53 1 ý 'jEa2ýNwl.

PVRC Low-Alloy Steel Component Tests (Environmental Shift + Size/Roughness Effect)C. E CD 4 (/0 100 90 80 70 60 50 40 30 20 10 0 1.E+03 1.E+04 Number of Cycles 091802 ppt 56 x N A-302 Failure X Failure A A-302 Initiation

---ASME LAS Mean Air -ASME Design 1.E+05 1.E+06 Flow Rate Effects Laboratory Data Versus Operating Experience

"* Absence of flow and uniformity of strain at low strain rate appears to be the major difference

"* Laboratory data are obtained at high strain amplitudes and very low strain rates on cylindrically-shaped (uniform surface strain) test specimens, under simulated reactor water flow chemistries and very low flow velocities.

"* Fatigue crack initiation mechanism appears to be rupture of the protective oxidation/passivation layer at high strain, with reoxidation/repassivation prevented by sustained straining at very low strain rates and low oxidizing potential.

091802.ppt 57 Flow Rate Effects Laboratory Data Versus Operating Experience Full-Scale Component Tests on Carbon Steel with Flow at Temperature

"* KWU Tube Tests (Now Framatome ANP GmbH), Erlangen, Germany "* Strain-Induced Stress Corrosion Cracking of Non-Post Weld Heat Treated Cold-Formed Bends and Welded Joints in High-Temperature Water, E. Lenz and A. Liebert, BMFT 11 B 504/2, May 1986.

  • 1200 Liters/hr (20 Liters/min) recirculating flow rate at 2400C with controlled dissolved oxygen concentration.
  • 34-mm nominal diameter tubes with 3.6-mm wall thickness bent 1800.
  • Cross-sectional area = 1.407 in 2; maximum flow rate = 14.45 in/sec (0.4 m/sec).091802.ppt 58 KWU Carbon Steel Component Flow Tests (Variable Flow, Variable DO)0.55 0.50 0.45 0.40 0.35 1. 0.30 0.25 0.20 1.E+02 Cycles 091802 ppt 59* AA U -, 3-*== * "',,.E 8 ppm DO, low flow
  • 0.01 ppm DO, low flow 0 0.2 ppm DO, low flow A 0.2 ppm DO, moderate flow -......ASME CS Mean Air -ASME Design, 2880C 1.E+03 1.E+04 1.E+05 I I Operating Plant Failure Data
  • EPRI TR-110102, Nuclear Reactor Piping Failures at U. S. Commercial LWRs: 1961 -1997, Draft Report, July 1998
  • 4064 failure event records; about 2200 leaking events and about 1800 non-leaking events
  • Mostly IGSCC (1227 failures), erosion and flow assisted corrosion (1003 failures), and vibratory fatigue (475 failures) events
  • The 636 fatigue failures are apportioned into 475 vibratory fatigue failures, 120 thermal fatigue failures, and 13 corrosion fatigue failures; 28 failures had no second-level failure description 091802 ppt 60 Operating Plant Failure Data "* All of the corrosion fatigue events were detected within the first 13 years of plant operation (4 in BWRs and 9 in PWRs) "* Similarly, the majority of the thermal fatigue events were detected within the first ten or eleven years of operation, caused by unanticipated transients not included in the design basis "* Similarly, most vibratory fatigue events were detected within the first fifteen years of plant operation, with low event frequency since 1985 "* Systems affected by corrosion fatigue were component cooling water (2), feedwater (2), RCS (2), core spray (1), safety injection (1), service water (1), small instrument line (2), steam line (1), and containment cooling (1)091802.ppt 61 Operating Plant Failure Data Result
  • Available U. S. failure data for nuclear power plant piping components does not support the observations that would be reached from the conservative (trickle flow) laboratory data there is no trend of accelerating corrosion fatigue failures in U. S. operating nuclear power plants "* In general, the laboratory fatigue data under simulated reactor water environmental conditions fall within the margins already accounted for by a portion of the ASME Code factor of 20 "* The effect of reactor water flow rate further mitigates against potential environmental fatigue failures "* The ASME Code fatigue design procedures and industry practice with respect to design-basis transient definitions are sufficiently conservative to more than compensate for reactor water environmental effects 091802.ppt 62 Data Evaluation Conclusion e Data are consistent with and support the revised probabilities of through-wall cracking (and leakage) for component locations evaluated in NUREG/CR-6674

"* Laboratory test data obtained under simulated reactor water environmental conditions show that almost all data for carbon and low-alloy steels are within margins in ASME Code fatigue design curve "* Structural/component fatigue test data with one surface in contact with water environment showed behavior accounted for by margins in ASME Code fatigue design curve "* Operating experience has not shown significant increases in fatigue failures ascribed to reactor water environmental effects as a function of increasing length of service of nuclear power plant components 091802.ppt 63 Aging Management of Environmental Fatigue for Carbon/Low-Alloy Steels Wrap-up Michael R. Robinson Duke Energy September 18, 2002 NRC Headquarters Enclosure 6 091802 ppt 65 Summary "* NUREGICR-6674 probabilistic calculations showed that, even with essentially bounding assumptions, the number of carbon/low-alloy steel component locations at which fatigue cracks would initiate and grow to a significant size are very few "* Industry recalculations of these probabilities, based on more realistic assumptions, show two to five orders of magnitude reduction of crack initiation and through-wall cracking probabilities

"* U. S. nuclear power plant piping failure data do not exhibit a clear trend supporting environmental fatigue concerns "* Review of laboratory and component/structural test data show that no explicit treatment of reactor water environmental effects is needed for carbon/low-alloy steel components 091802 ppt 66 Conclusions

"* MRP evaluation concludes consideration of environmental fatigue effects for carbon and low alloy steel components, as stipulated in GSI-190 closeout memorandum, is not warranted

"* All carbon/low-alloy steel fatigue locations can continue to rely on existing plant programs to track component fatigue usage through the license renewal period and remain in compliance with all NRC regulatory requirements

"* MRP evaluation of austenitic stainless steel locations is continuing 091802.ppt 67 Actions "* MRP will provide draft Interim Staff Guidance "* ISG to include technical basis "* Re-analysis of NUREG/CR-6674 (MRP-74) "* Environmental data review (MRP-49) "* ISG to include changes to GALL and SRP "* ISG to be provided by December 31 "* Coordinate revision to near-term guidance provided in MRP-47

  • Deferral of RAI process pending ISG and ongoing austenitic stainless steel activities 091802.ppt 68
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