ML031040164
| ML031040164 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/05/2002 |
| From: | Mathews L Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| -nr, FOIA/PA-2003-0018 | |
| Download: ML031040164 (29) | |
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MRP Update to ACRS Materials Subcommittee June 5, 2002 Larry Mathews Southern Nuclear Chairman, MRP Alloy 600 Issue Task Group AmRS 4M12.1
.MRPP e MRP Presentations Alloy 600 ITG Status Mathews 15 min Alloy 600 Crack Growth Rate Hickling 45 min l
Probabilistic Fracture Mechanics Model Riccardella 45 min Collateral Damage Mathews 10 min Technical Assessment of DB Degradation White 30 min Mechanisms Industry Inspection Plan Lashley 60 mli ACRS85122 t
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Freedom of Information Act, exemrFLions 5 FOIA-L ° °3-o i L.
Crack growth rate for thick-section Alloy 600 material exposed to -PWR primary water John Hickling, EPRI for the MRP Alloy 600 Issue Task Group ACRS44W2.5 MRP1.i, MRP Crack Growth Rate Approach:
Overview
- Goal was to establish appropriate CGR guidelines for generic application to thick-section Alloy 600 base material under PWSCC conditions
- MRP panel of international experts on SCC (includes ANLINRC Research) was established August 2001 and has met several times to date
- Extensive consideration was given to the likely OD environment in the annulus between a leaking CRDM nozzle and the RPV head (prior to Davis Besse Incident)
- Relevant arguments remain valid today as long as leak rates are low (typically i I liter/h or 0.004 gpm)
- Plant experience has shown this to be the usual case ACRS 5J2.6
OD Annulus Environment Most likely environments
- Hydrogenated superheated steam, if pressure drop within SCC crack
- situation has been considered in detail for the case usually observed to date, i.e. low leak rates (c Il/h) and little or no wastage of LAS vessel head
- full evaluation has not been performed for Davis Besse type situation involving cavity formation and extensive wastage as a consequence of boric acid corrosion AM JSJ.9 OD Annulus Environment
- Consideration of oxygen/hydrogen effects common to all three possible environments:
- Oxygenated crevice environment highly unlikely because:
- Back diffusion of oxygen Is too low compared to counterflow of escaping steam (2 independent assessments based on molecular diffusion models were examined)
- Oxygen consumption by metal walls would further reduce concentration
- Presence of hydrogen from leaking water and diffusion through upper head results In a reducing environment
- Even If concentration of hydrogen was depleted by local boiling, coupling between LAS and Alloy 600 would keep electrochemical potential low
- Corrosion potential will be dose to NI/MO equilibrium, resulting In PWSCC susceptibility similar to normal primary water ACRStSO.1:
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- Possible environment #2: PWR primary water within normal specifications
- Main focus of subsequent CGR data evaluation by Expert Panel AMRI £113 OD Annulus Environment Possible environment #3: Concentrated PWR primary water. For low leak rates (< 1 I/h) as mostly observed to date:
- pHr between 4 and 9A based on MULTEQ calculations
- Actual pHT range expected to be narrower-due to precipitation of complex lithium-iron borates
- A French experiment simulating a leak detected such borate compounds and estimated that pHT of the liquid phase was between 7 and B
- A further French test involving slow concentration of a fixed volume of primary water showed no formation of caustic after conc. factor 103 (calculated PHT was - 4.5)
Cleaning practices followed during head assembly should minimize contamination by sulfates and chlorides and steam flushing will help to remove any residual Impurities AMCRS10C.1 a
OD Annulus Environment Possible environment #3: Concentrated POWR.
primary water (con.)
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Ohio State study shows no significant effcrnpHjOi PWSCC CGR between values of 5 andat 330 C For pi.g values between 7.5 aTndL.GR increases slightly.
bu fceeainfcorol rudts or pHT 9
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f on CGR should conservativl Ioe incertainties an the exact composition of the annulus chemistry for 4 < pHT c 9 More acid environments as a result of large leak rates and local cooling of head were NOT considered, but limited data (Berge et al., 1997) suggests that high chloride and oxygen levels are required for IGSCC of Alloy 600 to occur ACRS 502.1'?
OD Annulus Environment: results of Ohio State study on effect of pH I...
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MRP CGR database for Alloy 600:
screening of available data
- No attention was paid to numerous tests where no crack growth due to PWSCC was actually observed
- Result of data screening was elimination from further consideration of 203 CGR data points for one or more reasons (main reason individually documented in report)
- Consolidated database contains 158 data points for average CGR during each test (consistent with ASTM practice for measunng fatigue CGRs) plotted at a representative K value (ranged from 14.3 to 54.0 MPaqm)
- All were obtained in controlled primary water using fracture mechanics specimens under either constant load or constant displacement conditions
- Some tests under active load involved periodic unloading (considered to give a potential accelerating effect which is relatively small, at least for susceptible maIerals)
CARS WC 1--
d MRP CGR database for Alloy 600:
periodic unloading used in W tests I'.
ACRS 61502J22
Derivation of MRP CGR Curve
- Because of the known importance of material processing parameters on CGR, the initial evaluation was based on a heat-by-heat analysis of the screened database
- Insufficient data points were available from any single heat over a wide range of K values to determine the form of CGR dependence on stress intensity factor
- Shape of curve to be fitted was adopted from the Scott equation, originally developed (1 991 )-usingnspection data for axial cracks in thehFltransitions of SG tubes )
- This much larger database of CGR measurements Is considered to provide a more reliable indicator for the form of the CGR versus K dependence:
- da/dt = a(K-9)0 with Scott exponent 1 1.16 A2S 5102J5 Derivation of MRP CGR Curve:
examples of original results (2 labs)
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IA 2iOl Derivation of MRP CGR Curve Distribution describing CGR variability was then taken as thq log-normal fit to the ordered median ranking of toe-j value§HteQBrl e-2 aFts, usin rngro-sllikely estimator methodology AM-m-,,
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Derivation of MRP CGR Curve I
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-.,..i Comparison of MRP database with available plant CGR data
- Large uncertainties exist in reported values of CGRs from operating plants due to:
- uncertainties in ultrasonic measurements of crack size at two or more different times
- uncertainties in the estimates of K, which depend on estimates of residual stress
- uncertainties in the actual operating temperatures of CRDM nozzles in different plants and in different countries
Comparison of MRP database with available pant CGR data 0.9 a *DF fied CI i;'
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Comparison of MRP database with available plant CGR data
- Agreement with French field data is quite reasonable considering the uncertainties involved
- Supports the choice of the 75th percentile curve from the MRP distribution as representative of the rates expected for axial crack growth in CRDM nozzles
- In no case did the actual measured CGR In the through-wall direction exceeds mmlyr (0. 1 6 hiyor data from French plants of fundam entally Vesiing ouse design
- This figure was adopted In France, independent of nominal upper head temperature, to justify continued operation with axial cracks up to 11 mm (0.43 inches) deep for a one-year fuel cycle A-
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- For evaluation of (hypothetical) OD cracking above the J-groove weld, the M RP recommends that CGR values from the curve be multiplied by 2x to allow for uncertainty in exact composition of the external chemical environment
- A subgroup of the Expert Panel have revisted the relevant arguments in the light of the Davis Besse experience and found that they remain correct as long as leak rates are low (typically less than I liter/h or 0.004 gpm)
- Analysis would no longer be valid, however, If leak rates were sufficiently high to result in a large, local decrease in temperature and appreciable corrosion of low-alloy steel
- Limited data on SCC in concentrated boric acid solutions indicate that
- Alloy 600 Is very resistant to TGSCC (material design basis)
- high levels of oxygen and chloride are necessary for lntergranular cracking to occur at all
- effects are then worse at Intermediate temperatures, suggesting that mechanism Is different from PWSCC ACRSM US.42
Outline of Presentation
- Overview of Probabilistic Fracture Mechanics Methodology for RPV Top Head Nozzle Cracking
- PFM Analyses in support of MRP Inspection Plan
- Susceptibility Categories
- Inspection Types and Frequencies AcRS~,
I^024 Key Elements of RPV Head Nozzle PFM Analysis
- Probability of Leakage
- Weibull Model based on Experience to Date
- Incorporated into Monte Carlo Model
- Fracture mechanics modeling for Stress Intensity Factors
- Through-Wall Cracks
- Part Through Wall Cracks
- Stress Corrosion Crack Growth Statistics
- Effect of Inspections
- Inspection Interval
- Inspection Reliability AM V..,7 U
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Weibull Distributior$'used in PFM
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I SCC Crack Growth Data for Nozzle Material in Reactor Environment 0.7 0 6 0.3 0.2 0.1 0.0 1E-13 AMR V=4 0e S3.12 IE-a, PV raw Constat a at IWC (6179F) 5-I10 h%
CGR Initiation vs. Growth
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BAC Tests Simulating Nozzle Leakage EPRI Annulus Test Matrix GWF
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.10 ls 400 0.01 CRRUC LILAK BAC Tests Simulating Nozzle Leakage Typical Sectioned EPRI Test Specimen 1 ~..
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Status of Inspection Plan
- Inspection Plan and technical bases were presented to NRC staff on May 22
- Technical Bases documents will be provided to NRC in June 2002.
- Comments received In following areas
- Plan should address inspections for both wastage and nozzle ejection issues
- Timeframe for wastage development
- Leakage past tight interferences
- Policy issue of detecting degradation through leakage
- Address replacement head ACRS 6J.102
- 02 Purpose
- Provide guidance and the basis for a long-term management program for RPV Head penetrations.
- Preserve structural integrity thereby ensuring safe operation.
GL 88-05 program remains the primary defense against boic acid wastage.
Inspection frequencies have been conservatively established relative to the structural Integrity of the RPV Head.
- Provide a graduated approach to inspections to allow early detection of leakage or through-wall cracking prior to challenging structural integrity or significant wastage.
Structural Integrity is defined as maintaining an acceptably low probability of developing cracking that could lead to nozzle ejection.
AMR 6/=0.103 a
Degradation Progression (con tinued)
- Condition 2. As the crack widens and the minimum leak path flow area increases
- Flashing-induced erosion or FAC may initiate the material loss process
- Galvanic corrosion may be Important if cooling Is sufficient to allow liquid to exist over a significant height In the annulus
- These mechanisms could be expected to produce greater relative material loss deep in the annulus, consistent with Davis-Besse Nozzle #2 and the EPRI SAC leaking annulus tests
- Condition 3. As the leak rate increases and the wastage area grows from a small cavity to a large.
open cavity
- Aerated boric acid corrosion (up to 1-5 Inches per year) may occur I'
'. 15 Degradation Progression (continued)
The geometry of the Davis-Besse Nozzle #3 cavity may indicate that aerated BAC removing material from the top surface down toward the cladding replaced corrosion and/or erosion deep down in the annulus as the dominant degradation mode
- The slope of the walls of the cavity change with distance from the top head surface Heat transfer calculations show considerable local cooling of the head for the range of leak rates believed to apply to this noznce, indicating an aerated, concentrated liquid boric acid solution fm on the top head surface adjacent to this nozzle Laboratory tests and plant experience indicate relatively high corrosion rates for low alloy steel exposed to aerated, concentrated liquid boric acid solution in comparison to other material loss mechanisms
- Gravity-driven flow of this liquid film would tend to produce the observed oblong shape of the Nozzle #3 cavity A.,
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Risk Informed Basis Probabilistic fracture mechanic (PFM) analyses using a Monte-Carlo simulation algo hm
- Included experience-based time to leakage correlations
- used a Weibull model of plant Inspections to date,
- fracture mechanics analyses of various nozzle configurations containing axial and circumferential cracks, and
- MRP developed crack growth rate data for Alloy 600.
- Performed to determine the probability of leakage and failure versus time for a set of input parameters:
- head operating temperature.
- benchmarked against experience to date
- Sensitivity studies were performed for various:
- inspection types (visual or NDE) and
- inspection intervals.
ACRSWS2.106 Risk Based Susceptibility
- Moderate susceptibility boundary:
- The number of EDYs at which a plant reaches probability of one leaking nozzle = 20%
(approximately equal tothe probability of net section collapse (NSC i.e. nozzle ejection) 1 x 104
- High susceptibility boundary:
- The number of EDYs at which a plant reaches:
- probability of nozzle eection I x 1O-3 (approximately equal to the probability of one leaking
- consistent with NRC RG 1.174 guidance for change In Core Damage Frequency.
CRS 6We.107 i
CRDM/CEDM Head Penetration Flaw Acceptance Criteria
- Visual evaluation criteria
- EPRI Technical Report 1006899, Visual Examination for Leakage of PWR Reactor Head Penetrations on TOD of the RPV Head: Revision IMarch 2002.
- Non-visual evaluation criteria
- MRP and ASME Section XI Code are working to develop final criteria, and until those criteria are issued, NRC-proposed criteria may be used.
ACRS6SJU5;2 110 Inspection Schedule - Low Susceptibility For low susceptibility plants (< 10 EDY,:
- Perform a Bare Metal Visual (BMV) examination of 100% of the CRDM/CEDM penetrations once per 10 years, beginning no later than the third ISI interval.
- Or, perform NDE (i.e., non-visual examination) of 100% of the CRDM/CEDM penetrations and associated J-groove welds once per 10 years, beginning no later than the third ISI Interval.
ACRSW5M02.111 a.
Inspection Plan Plants with leak(s) or through wall cracks identified:
- Discovery Inspection Perform a non-visual examination of the CRDMWCEDM penetrations and associated J-groove welds to characterize the crack or leak identified.
- Indications are evaluated or repaired in accordance with flaw evaluation guidelines.
ACRS 6IS.tu 1
Plants with leak(s) or through wall cracks Expansion of Inspection (to be Implemented no later than next RFO)
- Perform NDE ( i.e., non-visual examination) of 100% of the CRDM/CEDM penetrations and associated J-groove welds.
- Indications are evaluated or repaired in accordance with flaw evaluation guidelines (Reference 4).
- Or, perform an evaluation to justify continued visual examination until the RVH component is removed from service.
- Or. perform NDE at a frequency to be determined such that the 3x safety margin of a hypothetical circumferential crack growing above the weld Is not exceeded prior to the next Inspection.
ACRS 6U2 115
Multiplier on CGR Distribution for Within-Heat Variability eel~
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Inspection Plan PFM Runs:
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Inspection Frequency Runs:
Probabilities of Detection
- Bare Metal Visual Inspections (BMV)
- Initial POD = 0.6
- POD for Subsequent Exams = 0.2 x Initial POD (when Leakage missed)
- POD = f(crack depth) per EPRI-TR-1 020741
- 80% Coverage Assumed 1Dimitrijevic. V. and Amrnmirato. F.. Use of Nondestructive Evaluation Data to Improve Analysis of Reactor Pressure Vessel Integrity.
- EPRI Report TR-102074. Yankee Atomic Electric Co. March 1993 ACRSW62.I7 6?
Inspection Plan Technical Basis:
Effect of NDE Inspection iO A=A &W~V."O a
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a6 Effect of Inspections upon Entering Moderate Category vem-ua ISSIJO.?
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Deterministic Crack Growth Analysis Results Time for initial Flaw Size of 30 Circumference to Grow to 1650 Temperature and 300° (EFPY)
F) lWestinghouse-Type Plant 165 3000 580 23.7 31.7 590 18.3 24.6 600 14.2 19.1 602 13.5 18.2 605 12.5 16.8 ACRS 6=1402?4 Deterministic Crack Growth Results Added to Susceptibility Category Plot
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Approach (continued)
- The leak rate also determines the amount of boric acid deposits that exit the pressure boundary
- The results of corrosion and erosion rate evaluations are used to bound:
- The timeframe for significant degradation
- The volume of low alloy steel material loss versus the volume of deposits produced ACRS L=12.S6 Material Loss Mechanisms
- Corrosion mechanisms
- Erosion mechanisms
- Flow accelerated corrosion ACR
Material Loss Mechanisms Matrix PREUMINR Extmet *fWastae ktitia Tight Eabrted ca Carz Aimt..I Am,,itrs iDcaead Boic Acid Corrosion DryEA Or Blorie Oxidc Cryst Corosion C_..
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Flob Accrleried Corroin (PAC Possible if li4uid vwdockishih uikd y ItS.4k.m bthUi t
ph SwfwtOwn and wupenrgwyme bow mugh Oexn tasbilizes lrpingernew I Fbishing-aduced Erosim pFvibe ffdpitu fight taze ad mwamenam l4a dauck ln Op MC buuI
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C e Cosin leed o be ikely because low aleoy sites aI w
khomm IL!r kw wiP mmltp Ikd So" act passivte h an acraed coucet boric ad chaded Itua' Waahtic CeosIm =C biws *stne 14uid sohtion exioi
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t -e lo Acid Cuoiso aim. z ewit~~~ai3 h~i PossIel but we upected to bhe lout m Ibr aeted I lcied bric Acid Carmion liACI
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Degradation Progression Aef#W021,
Collateral Damage ACRs 482.r MRPi' Collateral Damage
- Indicated impact on Conditional Core Damage Probability should be insignificant
- No impact on ECCS capabilities
- Effect on shutdown reactivity capabilities minimal
- Impact and jet loads should not affect significant number of rods
- Loose parts also have only limited Impact AMAS
Contents
- Purpose and Approach
- Material Loss Mechanisms Corrosion mechanisms Erosion mechanisms
- Degradation Progression
- Boric Acid Corrosion Tests Simulating Nozzle Leakage NOTE: Additional infonnation and results ar provided in the May22. 2002. presentation to the NRC staff on this subject. wRich is available an te NRC website area for reactor head degradation.
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,V Purpose and Approach ACRS 6I02Je3
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