ML20215M672

Summary of 861001 & 15 Meetings W/Util,Bechtel & Westinghouse Re Issues Concerning Pipe Break Postulation Technology.Meeting Agenda & Viewgraphs Encl

ML20215M672
Person / Time
Site:	South Texas
Issue date:	10/22/1986
From:	Kadambi N Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8611030155
Download: ML20215M672 (49)

Text

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1. 'o UNITED STATES

'g E' 3. 'g NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 g E

\...../ OCT 22 ISE Docket Nos.: 50-498 and 50-499 APPLICANT: Houston Lighting and Power Company FACILITY: South Texas Project, Units I and 2

SUMMARY

OF MEETINGS ON PIPE BREAK POSTULATION TECHNOLOGY APPLIED TO THE SOUTH TEXAS PROJECT - OCTOBER 1, 1986 AND OCTORER 15, 1986.

The meetings were held to discuss a series of issues related to pipe break postulation at South Texas. Enclosure 1 shows the attendees at each of the meetings. Due to the large number of people present at the October 15, 1086, meeting and because of the unavailability of an appropriate conference room within the NRC staff's office space, the meeting was held at the Westinghouse offices. The agendas for the meetings are shown in Enclosure P. The information presented was a combination of proprietary and non-proprietory information. Enclosure 3 includes the non-proprietary slides used in the presentations. The proprietary information is being withheld pending submittal on the docket in a manner consistent with 10 CFR 2.790.

DISCUSSION:

The areas of discussion could be divided into three main categories, namely:

(1) the GDC-4 exemption request related to the pressurizer surge line (2) the request for relief from the acceptance criterion in SRP 3.6.?

related to the Cumulative Usage Factor (31 Other pipe break postulation methods which have been proposed for South Texas in FSAR amendment submittals.

The discussions in each category is briefly summarized below.

(1) Pressurizer Surge Line Exemption Reouest: Subseauent to receipt of the exemption request the staff had issued a" request for additional information. The applicant described the responses they proposed to make and requested staff feedback. The discussion at the October 1, 1986, meeting resolved some of the questions but also generated other requests for information. The applicant undertook to provide the needed informa'c ion on an expedited basis for the meeting on October 15, 1986.

The information presented addressed the staff's concerns with three areas requiring further resolution.

8611030155 861022 PDR ADOCK 05000498 A PDR

(a) Provide additional basis to justify the location chosen for the LBB evaluation in terms of actual stress values and material properties.

(b) The staff requested benchmarking of the leakage predictions against the available data from the Duane Arnold pipe cracking incident.

i (c) The staff requested some detailed explanation of analyses presented in WCAP-10489. The explanation is scheduled to be provided to the staff's contractor at a brief meeting with Westinghouse.

The staff indicated after the meetings that, subject to review of the information to be submitted formally on the South Texas docket, the technical resolution of the exemption request is at hand.

(2) SRP Deviation on CUF: The applicant has requested a plant specific deviation from the guidance in SRP Section 3.6.2 on CUF, Information had been submitted in a September 17, 1986, letter providing technical J

justification for the deviation at 57 specific postulated break locations. Additional information and justification was provided at the i neetings on October 1 and October 15, 1986. The staff requested further documentation of information on crack growth and the influence of vibratory loads.

Specifically, the staff requested the following additional documentation:

a) Discuss how the fatigue crack growth (FGC) analysis to be performed conservatively bounds the cases of CUF 0.1 covered by this request, b) Provide documentation on the FCG rate curve and its conservatism relative to PWR environment, c) Include in the FCG analysis the crack growth that would result from vibration loading of that line. Provide information to demonstrate the correlation between the stress levels reflected in the FCG analysis and i the stress levels permitted as a result of the piping walkdown,

! d) Provide data to justify the selection of the K threshold used in the FCG analysis and show how the results would be affected by_the use of a more

conservative value of K threshold such as 4 or 5 ksi/in, e) Document the use of ASME Section III C stress indices in the FCG analysis.

f) Discuss the service experience of the Class 1 Ifnes in the scope of this

!. request relative to vibration failures, in particular for the CVCS.

Subject to satisfactory docketing of the information, the staff felt that a

! favorable response to the applicant's request is indicated.

P i

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3. Other Pipe Break issues: The applicant has made submittals on each of the. issues, some of which have been addressed in a supplemental SER and others are under staff review. The applicant was specifically interested in the acceptability of the coupled stress analysis. The staff indicated that the proposed approach is acceptable.

i N. Prasad Kadambi, Project Manager PWR Project Directorate No. 5 Division of PWR Licensing-A

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, 3. Other Pipe Rreak issues: The applicant has made submittals on each of the issues, some of which have been addressed in a supplemental SER and others are under staff review. The applicant was specifically interested in the acceptability of the coupled stress analysis. The staff indicated that the proposed approach is acceptable.

! Al. #

l N. Prasad Kadambi, Project Manager PWR Project Directorate No. 5 Division of PWR Licensing-A h

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Mr. J. P. Goldberg Houston I.ichtino and Power Company South Texas Project-cc:

Brian Berwick, Esq. Resident Inspector / South Texas Assistant Attorney General Project Environmental Protection Division c/o U.S. Nuclear Regulatory Commission P. O. Box 12548 P. O. Box 910 Capitol Station Bay City. Texas 77414 Austin, Texas 78711 Mr. Jonathan Davis Mr. J. T. Westermeir Assistant City Attorney Manager, South Texas Pro.iect City of Austin Houston I.ighting and Power Company P. O. Box 1088 P. O. Box 1700 Austin, Texas 78767 Pouston, Texas 77001 Ms. Pat Coy Mr. P. I., Peterson Citizens Concerned About Nuclear Mr. G. Pokorny Power City of Austin 5106 Casa Oro P. O. Box 1088 San Antonio, Texas 78233 (

Austin, Texas 78767 Mr. Mark R. Wisenberg Mr. J. B. Poston Manager, Nuclear 1.icensing Mr. A. Von Rosenberg Houston I.ighting and Power Company City Public Service Boad P. O. Box 1700 l P. O. Box 1771 Fouston, Texas 77001  !

San Antonio, Texas 78296 Mr. Charles Palligan Jack R. Newman, Eso. Mr. Burton I. I.ex Newman & Poltziroer, P.C. Bechtel Corporation 1615 1. Street, NW P. O. Box 2166 Washingtor, D.C. 20036 Pouston, Texas 77001 Melbert Schwartz, Jr., Eso. Mr. E. R. Brooks Paker & Botts Mr. P. l.. Ranoe One Shell Plaza Central Power and I.ight Company Houston, Texas 77002 P. O. Box 2122 Corpus Christi, Texas 78403 Mrs. Peggy Buchorn Executive Director Citizens for Eauitable Utilities, Inc.

Route 1. Box 1684 Brazoria, Texas 77422

Houston lighting & Power Company South Texas Project '

cc:

Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Mr. I.anny Sinkin, Counsel for Intervenor Citizens Concerned about Nuclear Power, Inc.

Christic Institute 1324 North Capitol Street Washington, D.C. 20002 licensing Representative Houston I.ighting and Power Company Suite 1309 7910 Woodmont Avenue Bethesda, Maryland 20814 4

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Meeting Summary Distribution

.lDockef oFeeMrel- File NRC Participants NRC PDR Local PDR N. P. Kadambi PD#5 Reading File G. Bagchi J. Partlow H. L. Brammer V. Noonan R. Goel Project Manager Pei-Ying Chen OGC-Bethesda Ted Sullivan E. Jordan Barry Elliot B. Grimes Richard E. Johnson ACRS (10) Michael Mayfield M. Rushbrook Ron Ballard S. Lee Bob Bosnak cc: Licensee and Plant Service List

e SOUTH TEXAS MEETING ON PIPE BREAK POSTULAT!0N TECHNOLOGY OCTOBER 1, 1986 NAME AFFILIATION N. P. Kadambi NRC/NRR A. B. Poole HLA P Eng.

W. W. Watson Bechtel R. H. Reck Bechtel (Proj. Engr.)

S. Lee NRC/NRR/PWR-A Ed R. Johnson W - Engrg.

Seth A. Swamy W - Fracture Mechanics F. J. Witt W - Fracture Mechanics G. Bagchi NRC/NRR/PWR-A/EB H. L. Brammer NRC/NRR/PWR-A/EB R. Goel NRC/NRR/PWR-A/PSR Pei-Yino Chen NRR/PWR-B/ISAPD Warren Bamford W Structural Materials Energ.

Jack Bailey RL&P PE Lic Randall L. Beverly HLAP Le-PSI /ISI Ron Gamble NOVETECH Ted Sullivan NRC/PWR-A/EB Barry Elliot NRC/PWR-A/EB Richard E. Johnson NRC/DSR0/EIR Hector Diaz SWRf/NDE Level III Riyad Oashu Bechtel David H. Roarty W - Engin.

Mark R. Wisenburg RL& P - Deputy Proj. Mgr.

F. Paul Chen ETEC Alex Paterson W STP Licensing Harry Clark W Licensing D. F. Landers Teledyne Eng. Services

s SOUTHTEXASMEETING(CONTINUATION)

ON PIPE BREAK POSTULAT!0N TECHNOLOGY OCTOBER 15, 1986 NAME AFFILIATION N. P. Kadambi NRC/NRR/PD#5 G. Bagchi NRC/NRR/PAER P. L. Brammer NRC/NRR/PAEB J. J. McInerney W Licensing WR Speizaletti V Licensing Warren Bamford W Licensino Alex Paterson W STP Licensing Team Seth A. Swaymy W GTSD F. Joel Witt U GTSD J. T. Westermeirer RLAP Project Manager W. Paul Chen ETEC Ted Sullivan NRC/NRR/PAEB Ron Gamble NOVETECH Mark Wisenburg HLAP Deputy Proj. Mgr.

William Watson Rechtel Bruce Poole HL&P Eng.

Michael Mayfield NRC/RES/ DES Harry Clark W Licensing Pei-Ying Chen NRC/PWR-B/ISAPD Ron Beck Bechtel Ron Ballard NRC/PWR-A/EB Maurice Axelrad NAH (PLAP)

W. F. Anderson  !&E, NRC A. 7accaria Bechtel Proj Mar.

Jack Bailey HL&P Lic E. Johnson Westinghouse D. H. Roarty Westinghouse R. Elliot NRC/PWR-A/ER S. Lee NRC/PWR-A/EB Bob Bosnak NRC/DSR0

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W /NRC Meetlag on Pipe Break Issues October 1, 1986 Agenoa 8:30 - 8:35 o In:roduction M. R. Wisenourg (HL&P) 8:35 - 8:45 o Description of Pipe-Whip fustraints at STP A. B. Poole (H.&P) 8:45 - 10:45 o General Discussion of Responses S. Swamy (Westinghouse) to P.C Q's on Pressurizer Surge 1.ine Leak-Before-Break o

, Discuscion of H.&P Submittal on BOP A. B. Poole (H.&P) leak-before-break o 01scussio1 of STP FSAR cnanges far deletion of RCS Main Loop and Cross- 8. Wetton (Bechtel) over piping breaks in accordance with nar ow scopa modification to GDC-4 o

NRC caucus and feeoback regarding' leak-before-brtek issues 10:45 - 11:45 o Surtriary of major changes to STF 8. Watson (Bechtet)

FSAR Section 3.6 regarding 100 jet criteria, coupled otress analysis (terminal and vs. branch connections),

clarification of plastic hinge criteria, whip on corvjult and Lse of non Category 1 eqaipment o mC caucus and feedback regarding miscellaneous pipe break issues 11:45 - 1:15 o LimCH 1:15 - 3:15 o Su'inary Presentation on CUF (H.5?

Technical Report) A. B. Poole (HL&P) o Response to MC Q's on H.&P Tec*.nical Repert on CW o NRC caucus and' feedback regarding cur l

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AGENDA

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SOUTH TEXAS PROJECT PIPE BREAK ELIMINATION OCTOBER 15,'1986 e SURGE LINE DISCUSSION i . BACKGROUND

. ANSWERS TO OCT. 1, 1986 QUESTIONS ON MATERIAL, f

STABILITY ANALYSIS AND NUREG 1061 ',

. J-INTEGRAL DOCUMENTATION

, LEAK RATE CALCULATION BENCH MARKING s ACCUMULATOR LINE DISCUSSION

. CRACK SIZE DETERMINATION - LEAK RATE

. MARGIN ON LEAKAGE SIZE CRACK e GENERAL DISCUSSION

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ELIMINATION OF REACTOR COOLANT LOOP RUPTURES

• Limited scope change to 10CFR50, GDC 4

• Applies to cold leg, hot leg and crossover leg

^ Ovnamic effects of RCL ruptures eliminated

^ Non-mechanistic ruptures retained in 3 areas e WCAP-8691 (Proprietary)/WCAP-8692 (Non-proprietary)

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1 DYNAMIC EFFECTS l

• Pipe whip interactions i e Break reaction forces e Jetimpingementforces e Missile generation j e Decompression waves within ruptured RCL l e Pressurization in cavity, subcompartments and ,

! compartments I

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• Break propagation criteria l

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DESIGN FEATURES AFFECTED 1 l l e RV cavity HVAC blowout panels - Eliminated ,

e Class I stress analysis - Based on RCL branch line ruptures a Heavy components / supports - Based on RCL branch line ruptures

^ Subcompartments- Reduced design pressures i

scs6(ts.m 33

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i NON-MECHANISTIC RUPTURES

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No change in FSAR/ design bases

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Emergency core cooling system i

Containment pressure boundary

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e Equipment qualification - Environmental conditions l

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PIPE BREAK CR!TERIA ..

10 DIAMETER JET LIMITATION e Limited to high pressure - 870 to 2465 psia e Subcooled orflashing liquid e Unprotected components within 10 diameters assumed to fail

• Unprotected component beyond 10 diameters assumed undamaged

f PIPE BREAK CRITERIA USE OF NON-SEISMIC CATEGORY I SYSTEMS e To mitigate consequences of pipe break other than main steam system pipe break ,

e Subject to power availability

• Equipment must be qualified for break environment l

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TERMINAL END CRITERIA  ;

o BRANCH TECHNICAL POSITION MEB 3-1 "A BRANCH CONNECTION TO A MAIN PIPING RUN IS A TERMINAL END OF THE BRANCH RUN, EXCEPT WHERE THE BRANCH RUN IS CLASSIFIED AS PART OF A MAIN RUN IN THE STRESS ANALYSIS AND IS SHOWN TO HAVE A SIGNIFICANT EFFECT ON THE MAIN RUN BEHAVIOR."

o CURRENT STP FSAR PARAGRAPH 3.6.2.1.1.1.b.l(c)

" BRANCH INTERSECTION POINTS ARE CONSIDERED A TERMINAL END FOR THE BRANCH LINE EXCEPT WHERE THE BRANCH AND THE MAIN P! PING SYSTEMS ARE MODELED IN THE SAME PIPING STRESS AflALYSIS Ati0 THE BRANCH LillE IS SHOWN TO HAVE A S!'2N!FICANT EFFECT ON THE 11AIN RUN BEHAVIOR (i.e.. THE N0l11NAL SIZE OF THE BRANCH LINE IS AT LEAST ONE HALF 0F THAT OF THE MAIN OR THE RATIO 0F THE It0 MENT OF INERTIA 0F MAIN RUN P!PE TO THE BRANCH LINE IS LESS THAN ,

10)."

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TERMINAL ENO CRITERIA o PROPOSED CHANGE TO STP FSAR PARAGRAPH 3.6.2.1.1.1.b.l(c)

" TWELVE INCH (12") AND LARGER PIPING CONNECTED TO THE RCL MAY BE MODELED WITH THE RCL IN THE SAME P! PING ANALYSIS AND, THEREFORE, CONSIDERED A PART OF THE l'AI'l RUN. OTHER BRANCH INTERSECTION PO!f4TS ARE CONSIDERED A TEFMIflAL END FCR THE BRANCH LINE EXCEPT (1) WHERE THE BRANCH AND THE MAIN PIPlflG SYSTEMS ARE MODELEO IN THE SAME P! PING STRESS ANALYSIS AND THE BRANCH LINE IS SHOWN TO HAVE A SIGNIFICANT EFFECT ON THE MAIN RUN BEHAVIOR (l.e., THE NOMINAL SIZE OF THE BRANCH LINE IS AT LEAST ONE HALF OF THAT OF THE MAIN OR THE RATIO OF THE MOMENT OF INERTIA 0F MAIN RUN P!PE TO THE E?A'iCH LINE IS LESS THAN 10), OR (2) WHERE, REGAROLESS OF S!?E CR "C"ENT OF INERT!A RATIO, THE BRAflCH LINES ARE SHORT IN LENGTH

0 4.'E

, '.0 5::'.'FICANT RESTRAINT DUE TO THERMAL EXPAfiSION."

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l PIPE BREAK CRITERIA e Pipe whip on rigid steel conduit Equal / greater diameter and wall thickness undamaged Smaller diameter assumed to be damaged and cables fail l

• P'.,s"c hinge / pipe whip No at plastic hinge point I ' 'opped by structure / component when shown to be j su fficiently strong enough I -

Plastic hinge point fomulae clarified e Protect only safety related instrumentation required to '

mitigate rupture

• Longitudinal break criteria changed to agree with BTP MEB 3-1 e Isolation restraints no longer required to protect check valves f

t Application of Leak-Before-Break to i

Balance of Plant Bruce Poole Houston Lighting & Power l

List of High Energy Lines for LBB on South Texas Project Auxiliary Lines Stainless Steel Material

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• 12" RHR Lines

• 12" Accumulator injection Lines l

• 10"/8" Branches off Accumulator Injection Lines

] e 8" Silines off RC hot legs l

• 4" Pressurizer Spray Lines from the Loop Connections 2

to the-Spray Valves e 4" Charging, Alternate Charging, and Letdown Lines '

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• 3"/6" Auxiliary Spray, Loop Drain, and Excess Letdown Lines I

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l List of High Energy Lines for LBB on i South Texas Project Auxiliary Lines Carbon Steel Material

• 30" Main Steam Lines e 16"/18" Main Feedwater Lines

• 8" Auxiliary Feedwater Lines e 4" Feed Bypass Lines

• 3"/4"/6" Steam Generator Blowdown Lines 1 .

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l l Screening Criteria Parameters l

• Piping stress review

-Define required data packages

-Develop procedure and flow charts

-Obtain thermal transients and cycles

{ - Obtain moment loadings I

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l Screening Criteria Parameters i

• Piping stress review (cont)

-Summarize ASME equations 12 and 13

stresses for Class 1 systems i

j -Summarize cumulative usage factor for Class 1 systems

- Calculate-transient stresses for non-Class 1 systems - .

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l 1-i l I l Screening Criteria Para ~ meters Results of Screening

• Group 1 i

- Lines which pass limit moment screening and i no additional evaluation is required '

• Group 2

- Lines which pass longitudinal stress screening i

and additional finite element analysis for crabk stability may be required

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1 l Screening Criteria Parameters i

l Leak-Before-Break l Evaluation I

I Flaw Selection

! Fatigue Crack Mechanistic Pipe j Growth Evatuation Break Evaluation i

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Leak Rate i

i Predictions 4

j Leak-Before-Break Screening

Criteria for Fatigue Crack Growth

) FCG Screening I

Review System Transients and identify Worst Systems

! for Fatigue Crack Growth I

Perform FCG Evaluation

) on Selected Systems l 1 Review i Prepare Report Conservatism is Crack Growth on Crack Stability, Acceptable? in Loadings or Leak Detection Through-Wall i and FCG Stresses and Yes No Re-evaluate i

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4 Screening Criteria Parameters

• Results of piping stress review

- Summary of potential intermediate break locations

-Summary of moment loadings

-Summary of mechanical properties

- Summar of transient stresses

- Recommendation of. cases for minimization of' postulated pipe break impact study

Summary

• The technology has been accepted for eliminating I PWR reactor coolant loop breaks -

• The same technology has been applied to other 4

piping i

e The leak-before-break screening process identifies cases for which subsequent detailed analysis will successfully demonstrate L-B-B e More rigorous analysis will minimize the number of high stress breaks

• The number of pipe whip restraints and shields are minimized

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BACKGROUND e NRC letter to ANS 58.2 working group /provides draft October 1,1985 justification for revising the CUF from 0.1 to 0.4 e ANSI /ANS 58.2 d. aft revision 3/ includes revising CUF December,1985 i

from 0.1 to 0.4

• ANSI working group meeting /those present or February 12,1986 reporting agreed to raise the threshold CUF for postulating ruptures from 0.1 to 0.4 and include in next revision of ANS 58.2 e HL&P issues letter ST-HL-AE-1611 to the NRC/ states February 28,1986 that CUF of 0.4 has been incorporated into STP i procedures and is being used and to be included in future FSAR admendment e NRC issues NUREG-0781 providing safety evaluation April,1986 report for South Texas Project /SER references that CUF of 0.1 is being used at STP.

e HL&P issues letter ST-HL-AE-1722/ identifies that the August 14,1986 revised CUF criteria has eliminated approximately 50 postulated breaks and that facility response analyses have not been performed for these 50 points.

• HL&P issues letter ST-HL-AE-1758/providing September 17,1986 additional justification. -

HELBA ACTIVITIES AT STP i SINCE FEBRUARY,1986 l

l e The Unit 1 design work and facility response analysis work has been done in the following order of priority: ,

j 1. Terminal end break postulation points

2. Points with CUF >0.4

3. Points with equation 10,12, or 13 stress intensity range exceeding

. 2.4 SM

4. Points with CUF between 0.1 and 0.4 i

• Transmitted letter to NRC on use of LBB on pressurizer surgeline e Transmitted letter to NRC on program for extending LBB to BOP piping inside containment for Unit 2 only r Transmitted letter to NRC providing for use of 10 diametcr screening criteria for certain piping systems I

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i INTRODUCTION

e During the design, construction, and testing of a PWR, significant work is involved in reviewing the effects associatec' with postulated rupture of high energy piping.

e SRP Section 3.6.2 states that breaks need to be postulated at every location where the fatigue cumulative usage factor (CUF) is greater than 0.1 and/or equation 10,12, or 13 stress intensity range. exceeds 2.4 SM.

a Based upon the South Texas Project piping analysis completed by Bechtel, there are seventy-seven (77) points where CUF is larger than 0.1.

o Of these points, fifty-seven (57) have a CUF between 0.1 and 0.4.

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• Determining if protection at these aoints is needed and providing such l

protection for pipe whipping and c ischarging fluids is a significant effort at STP.

i e Reducing the number of these points will eliminate approximately 20 pipe whip restraints and jet impingement barriers from being installed at.

STP Unit 1. This will result in reduced worker radiation exposures and improved effectiveness of in-service inspections.

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I STP PROPOSED USE OF CUF = 0.4 l

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Only on Class 1 piping-

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All piping is austenitic stainless steel a

All piping systems are operated with fluid meeting PWR .

l water chemistry controls (oxygen < 0.005 PPM).

i l e All welds involved are butt welded joints.

I i e AI piping systems have been carefully designed, fabricated, and inspected.

1 e All piping systsms have had welded joints examined during l pre-service inspection program, no defects found.

i i e Leak before break (LBB) has been demonstrated on surge-i . line and all other lines are being evaluated for LBB to eliminate all postulated breaks for Unit 2. l .

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STP POSITION RELATIVE TO KNOWN l

l REGULATORY AND INDUSTRY CONCERNS

• ASME fatigue curves are not conservative

) STP - Based on 400 reactor years of operation and crack growth studies in

PWR environment, ASME curve for austenitic stainless steel is

! conservative for PWR design.

i e Environmental effects could cause earlier crack initiation and increased crack growth rates over tested ASME samples.

STP - For PWR chemistry and austenitic stainless steels, crack growth data shows improvement on threshold and no significant impact on crack growth.

• Small cracks could be present in piping prior to start of operations and CUF would not take this into account.

STP - Plant fabrication and in-process inspection provide high integrity joints

- Pre-service NDE inspections have been completed on each weld.

- LBB and crack growth studies show that even if a dery small defect is present, a margin of approximately 10 plant lifetimes exist at a point ,

with CUF = 0.4. .

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SERVICE EXPERIENCE -

STAINLESS STEEL LINES

• Over 400 reactor years of operation

! e No service-induced cracking of Class 1 stainless PWR lines a Primary coolant water environment is low oxygen

< 0.005 PPM, and very carefully controlled .

• Only service-induced cracking of any stainless lines was with a different environment

- BWR lines (oxygen)

. - Stagnant boric acid transfer lines (not Class 1)

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EFFECTS OF WATER ENVIRONMENTS ON

CRACK GROWTH IN CARBON STEELS l a ASME fatigue crack growth reference law (carbon l steel) shows a large effect of the environment

! e BWR environmental tests of SA 333 GR 6 carbon steel

! piping show that the S-N curve contains very little, if

any, conservatism

! e These test results are consistent with the ASME crack growth reference curves

)

• ASME sub-group on fatigue strength is considering revision of the carbon steel design curve because of this finding  ;

1 I PWR ENVIRONMENTAL EFFECTS ON STAINLESS STEEL Fatique threshold stress intensity factor j e PWR environmental effect is to increase the threshold stress intensity factor for crack growth i TP 304 at room temperature (USAMI):

l KTs = 2.8 KSI s/iN Dry Air

! K7g = 5.0 KSI sfiN Wet Air l TP 304 at 472"F (USAMI):

Wet Air l Krn = 6.0 KSI s/iN ,

TP 403 in 600 F steam (LIAU): ,

KTn increases by approximately 2 KSI s/iN vs Dry Air I

Estimated value for PWR environment

{ K1s = 8.0 KSI s/IN Fatique crack growth rate

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• Type 304 and 316 were studied in PWR environment along with associated welds  ;

l e Results showed very little influence of environment on crack growth I (Bamford, TRANS ASME 1979. - -

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i FATIGUE CRACK GROWTH FOR l TYPICAL PWR CLASS 1 PIPING Wall Thickness 1

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1.0 T i

i i Crack growth over one plant lifetime.

for and initially assumed flaw of == 0.1T 4

and == 0.15T. The Points used had

! usage factors = 0.6 i

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1 m2T 0.1 T l 1 1 I 10 20 30 40 n

Time (Years) One Plant Lifetime -

.ep Lew 13

i STP POSITION RELATIVE TO KNOWN

!. REGULATORY AND INDUSTRY CONCERNS (contd>

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• The potentially high stresses that might be associated with waterhammer

foads are not included in ASME fatigue calculations.

l STP - The lines in which postulated pipe ruptures would be eliminated have i been designed to minimize or preclude waterhammer effects.

I Safety injection i All lines are water solid at ambient temperature.

i l Chemical and volume control system 1

The high temperature lines have been designed to maintain water

. solid conditions during normal operation.

Reactor coolant system System designed to preclude steam void formation and waterhammer i not considered to be a concern.

l Residual heat removal system All lines in the system are designed to be water solid and waterhammer not considered to be a concern.

e Vibratory loads could exist that are not included in ASME fatigue j . calculations.

STP - During hot functional testing the Class 1 lines will belmonitored for steady state vibration. Vibration am plitudes will be related to stress levels and shown to be less than end urance limit. -

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INTEGRITY AND QUALITY OF CLASS 1 PIPING e Carefully designed, fabricated, and inspected e High energy Class 1 lines have demonstrated excellent fracture resistance-

- Type 304,304L

- Type 316,316L

- Associated welds e Leak before break (LBB) has been demonstrated generically for the main loop of Westinghouse plants, and specifically for South Texas (WCAP 10559).

• NRC sponsored studies at LLNL confirmed LBB. This led to revision of GDC-4 for the main loop.

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• Broad scope revision of GDC-4 for all high energy lines has now been l issued for comment.

Detailed analysis has been completed for South Texas s, urge line (WCAP 10489). ,

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JUSTIFICATION FOR l INCREASED USAGE FACTOR -

! Crack Depth /

Leak Before Break Thickness (a/t) Failure 1.0 O -

Tested stainless steel pipe through wall leackage k Leakage I\

[l CUF of 0.8 at CUF = 50.0 (average) 50.0 NUREG - CR 3243 Appendix C I 0.6 I

Worst case example f

, based on NUREG - CR 3059 1 0.4

! Detectable Q 0.2 j

~

f l /

l Initiationa 0 a a vn - - - - t/ a j CUF= 0.1 CU F = 0.4 CU F = 1.0 CU F =' 15.0

\

, Operating Lifetimes ,

ESTIMATES OF FAILURE LIFETIME Load Condition Estimated Lifetime Lifetime (i.e. CU F) to Wall Penetration to Failure 0.1 15/.1 = 150 Lifetimes 155 Lifetimes 0.4 15/.4 = 38 Lifetimes 43 Lifetimes Based upon actual stainless steel piping fatigue tests.

I e

VG F 6 1R1ep tew

l 1

1 l

j ,

! NUREG/CR-3982 CRACK GROWTH DATA I

i Years Required to Grow initial Flaw Size Flaw to Critical Size 2 percent (0.050 in.) 6,200 years '

l 8.6 percent (0.216 in.) 1,320 years l -

Point had CUF = 0.06929 i

l 1

l l

VG F 6 20 5ep tew

i MARGIN ON FAILURE section iii Usage Lifetimes ~~l/CUF j Factor ~CUF (1 = 40 Years) l 1.0 3 1.0

! 0.9 -

Best estimate for worst case Section ,

i Ill Fatigue Analysis using highest 0.8 -

possible stress concentration. -

1.2 0.7 -

0.6 - a W Point, CUF = 0.6 _

1,7 0.5 -

A NUREG/CR-3982 ,

CUF =.06929 '

l 0.4 --------

400 Years -

2.5 I

I l 0.3 -

i i

1 0.2 -

I 5.0 O.1 - - - - - -

l - - - - - - - - - - 1,072 Yea rs '

I 10.0 I . I I I i 1 -

0 0 5 10 15 20 25 ;30 35 Lifetimes (1 = 40 years) based upon time for crack growth ' ,

to critical flaw with initial flaw of == 10% through wall. .

<,,.s.,,,_,.

i COMPARISON OF LIFETIMES FOR NO DEFECTS AND DEFECTS PRESENT '

i Local Pipe Loading as Number of Years to Grow Quantified by Cumulative a Crack to the Critical Usage Factor Flaw Length No Initial Initial 0.1T j Defect Depth Defect

! CUF = 0.1 6,200 Years 1,072 Year.s l CUF = 0.4 1,720 Years 400 Years 1

I VG f n 19 5ep t ew

l

.I l

I

! IMPACT OF CHANGE IN I CUMULATIVE USAGE FACTOR l NUMBER OF BREAKS BY CUF RANGE i

l System ID No. of Breaks Line Size CUF Range l Normal Charging to Loop 1 3 4 .12 to .17 Alternate Charging - Loop 3 4 4 .14 to .26 i Aux. Press Spray 2 2 .12 to .19 l Excess Letdown loop D 3 2 .12 to .26 l Pressurizer Surge 14 16 .10 to .35  !

l Spray Line 2 6 .34 to .39 j Si injection Loop 1 11 12 .11 to .21 l SI injection loop 2 11 12 .11 to .16 Sl Injection Loop 3 7 12 .10 to .29 j No Socket Weld Breaks included Above j Total 47 .1 to .2 Total 5 I .2 to .3 Total 5 .3. to .4 .

Vf . I 8- 1 I '.e-gi t rw i

1 j

4

\ .

! CONCLUSIONS i

i l

• Criteria of using EU = .1 is too conservative 1 l

! e Fatigue curve not adversely affected by PWR l environment

! e Piping vibration will be closely monitored l e Only stainless steel pipe included i

i e Recommendation of a usage factor of.4 is reasonable i

l I

i 1 e* *