NL-12-2014, Proposed Inservice Inspection Alternative FNP-ISI-AL T-13 Version 1.0

Proposed Inservice Inspection Alternative FNP-ISI-AL T-13 Version 1.0

ML12276A110
Person / Time
Site:	Farley
Issue date:	10/01/2012
From:	Ajluni M Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-12-2014, FNP-ISI-ALT-13, Version 1.0
Download: ML12276A110 (64)

Text

Mark J. Ajluni. P.E. Southern Nuclear Nucl ear Licensing Director Operating Company. Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7 673 Fax 205992.7885 October 1, 2012 SOUTHERN'\'

COMPANY Docket Nos.: 50-348 NL-12-2014 50-364 U. S. Nuclear Regulatory Commission ATrN : Document Control Desk Washington , D. C . 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 & 2 Proposed Inservice Inspection Alternative FNP-ISI-AL T-13 Version 1.0 Ladies and Gentlemen:

In accordance with 10 CFR 50 .55a(a)(3)(i) , Southern Nuclear Operating Company (SNC) hereby requests Nuclear Regulatory Commission (NRC) approval of proposed inservice inspection (lSI) alternative FNP-ISI-ALT-13 Version 1.0. This alternative extends the inspection requirements of ASME Code Case N-770-1, Inspection Item B for Reactor Pressure Vessel (RPV) Cold Leg (CL) Dissimilar Metal (DM) welds from every second inspection period to once every 10-year interval for the Farley Nuclear Plant (FNP) Units 1 and 2. This alternative is based upon the technical basis given in ERPI document Materials Reliability Program (MRP) -349, which is provided as Enclosure 2.

The approval of FNP-ISI-ALT-13 Version 1.0 would permit the Unit 1 RPV CL volumetric examinations currently scheduled for fall of 2013 to be deferred until fall of 2017 (+/- 1 year), and would permit Unit 2 RPV CL volumetric examinations currently scheduled for spring of 2016 to be deferred until spring of 2020 (+/- 1 year). Exams would occur during or before the spring 2018 outage for Unit 1, and during or before the fall 2020 outage for Unit 2. Approval of this alternative is respectfully requested by May 28, 2013 to allow SNC sufficient time to schedule appropriately .

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.

Sincerely, M. J. Ajluni Nuclear Licensing Director MJAlRMJ/lac

U. S. Nuclear Regulatory Commission NL-2014 Page 2

Enclosures:

1. Proposed Alternative FNP-ISI-AL T -13 Version 1.0 in Accordance with 10 CFR SO.SSa(a)(3)(i)

2. Materials Reliability Program: PWR Reactor Coolant System Cold-Loop Dissimilar Metal Butt Weld Reexamination Interval Extension (MRP-349) cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. T. A. Lynch, Vice President - Farley Mr. B. L. Ivey, Vice President - Regulatory Affairs Mr. B. J. Adams, Vice President - Fleet Operations RTYPE: CFA04.0S4 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Farley Mr. E. L. Crowe, Senior Resident Inspector - Farley

Joseph M. Farley Nuclear Plant - Units 1 & 2 Proposed Inservice Inspection Alternative FNP-ISI-ALT-13 Version 1.0 Enclosure 1 Proposed Alternative FNP-ISI-ALT-13 Version 1.0 In Accordance with 10 CFR 50.55a(a)(3}(i}

Enclosure 1 Proposed Alternative FNp*ISI*ALT*13 Version 1.0 In Accordance with 10 50.55a(a)(3)(i)

Plant Site-Unit: eph M. Nuclear Plant (FNP) Units 1 & 2.

4th Inservice Inspection (lSI) Interval December 1,2007 through November Interval Dates:

30,2017.

Requested Date for Approval is by May 28,201 Approval:

ASMECode The affected components are Category Components N*770*1 Item specific components are provided in Affected:

Applicable edition and (for the 4th lSI interval) is ASME Code Edition Section XI, for Inservice Inspection of Nuclear Power Plant and Addenda: components," 2001 Edition through 2003 addenda.

10CFR50.55a (g)(6)(iiHF) requires ,,,"' ...',,,'"

Applicable pressurized*water reactors as of July Code of ASME N-77Q-1.

requires butt weld at < 580°F to be volumetrically examined every MRP-349 provides the overall basis for extension of the volumetric inspection interval for the Reactor re Vessel (RPV) Leg (CL)

Dissimilar Metal (OM) welds from every inspection as currently required by N-770-1, to a 10-year inspection Reason for that Request:

to 10 years while maintaining an level of quality Therefore, Southern Nuclear Operating Company (SNC) is requesting approval of this alternative to allow the use of the lSI interval extension for the affected FNP - Unit 1 and 2 components.

E1*1

Enclosure 1 Proposed Alternative FNP-ISI-ALT-13 Version 1.0 In Accordance with 10 CFR 50.55a(a)(3)(i)

SNC is requesting extension of the requirements of Code Case N-770-1, Inspection Item B for the RPV CL OM Welds from every second inspection period to once per 10 year inteNal.

Specifically, this proposed alternative would permit the deferral of the CL volumetric examinations currently scheduled for fall of 2013 for Unit 1 (baseline exams performed in fall of 2007) to be moved to the fall of 2017 (+/- 1-year as allowed by ASME IWA-2430 to allow inspections to coincide with the plant outage). Exams would occur during or before the spring 2018 outage. For Unit Proposed 2, this would allow examinations currently scheduled for the spring of 2016 Alternative: (baseline exams performed in the spring of 2010) to be moved to the spring of 2020 (+/- 1-year as allowed by ASME IWA-2430 to allow inspections to coincide with the plant outage). Exams would occur during or before the fall 2020 outage.

In addition to the volumetric exams performed to the specifications of ASNIE Appendix VIII, a supplemental eddy current test will be performed to the specifications of Code Case N-773. This matches the eddy current techniques that our vendor, Westinghouse, has previously used for qualification, coverage, and examination.

The overall basis used to demonstrate the acceptability of extending the inspection inteNals for Code Case N-770-1, Inspection Item B components is contained in MRP-349 (Reference 1, Enclosure 2). In summary, the basis for extending the inteNals from two lSI periods to 10 years is: (1) there has been no service experience with cracking found in RPV CL OM welds, (2) crack growth rates in RPV CL OM welds are small, and (3) likelihood of cracking or through wall leaks is very small in RPV CL OM welds.

SeNice Experience Basis for Use: Each unit's baseline exams were performed using remote mechanized examinations from the Inside Diameter (10) in accordance with Appendix VIII using performance demonstrated methods where 100% of the flaws were detected. The technique used in site specific exams included 100% coverage for axial and circumferential flaws. Data is obtained using encoded techniques; therefore, data may be reviewed by multiple qualified examiners.

Site specific mock-Ups were not used because of the flat, uniform surface associated with performance of these examinations from the 10. These techniques provide a strong assurance that flaws will be detected during inspections. Each FNP CL is exposed to approximately 550°F (CL Temperature) during normal plant operation.

All dissimilar metal welds in pipes 4" Nominal Pipe Size (NPS) and greater, including those containing Alloy 82/182, in ASME Section XI Category B-F, have been volumetrically examined every 10-years in accordance with ASME Section XI. There have been multiple instances in the industry in which E1-2

Enclosure 1 Proposed Alternative FNP-ISI-ALT-13 Version 1.0 In Accordance with 10 CFR 50.55a(a)(3)(i)

Primary Water Stress Corrosion Cracking (PWSCC) have in Alloy 82/1 nozzle-to-safe-end weld region of the outlet nozzle where temperatures range typically from to however, there are no known instances of PWSCC occurring in large bore (diameter greater than 14" NPS) that operate or near CL temperatures (approx. 550°F) within the nuclear industry. In summary, to date there have been no safety or structural integrity concern that has resulted from PWSCC in CL butt welds in the nuclear industry.

Crack Growth Rates (Flaw Tolerance)

All of the flaw tolerance analyses performed to date have shown that the critical crack in large-diameter butt welds operating at temperatures are very large. Assuming that a flaw initiates, the time required to grow to through-wall is in excess of 20 years in most cases analyzed. time to grow from a through-wall leak to a crack equal to the critical crack size can be in excess of 40 years.

More recent analyses have performed for the RPV through-wall residual distributions that were developed based on the Basis for Use: recent guidance. analyses have shown that the flaw tolerance of (Continued) locations is high and postulated circumferential flaws will not reach the maximum ASME allowable depth in less 10 years. Crack growth analysis is given for limiting plants part-circumferential through-wall flaws in of Enclosure Analyses have been performed to calculate the probability of failure for Alloy 82 welds using both probabilistic fracture mechanics and statistical methods. Both approaches have shown that likelihood of cracking or through-wall leaks, in large-diameter CL welds, is very small. Furthermore, sensitivity studies using probabilistic fracture have shown that even for more limiting high temperature locations, more frequent inspections than required by Section XI, as that in MRP-139 or Code N-770, have only a small benefit in terms of risk.

Though service experience may not an absolute indicator of the likelihood of future cracking, the experience does give an indication of the relative likelihood of cracking in CL temperature locations versus hot leg temperature locations. While there is a significant amount of PWSCC service experience in hot leg locations, the number of indications in butt welds is small relative to the number of potential locations. Also, all indications have detected they were a safety concern. Therefore,

! if hot leg PWSCC is a leading indicator for CL PWSCC, and the higher

'---_ _ _ _ _

• fre uenc of ins ections will be maintained the hot Ie locations, it is

Enclosure 1 Proposed Alternative FNP-ISI-ALT -13 Version 1.0 In Accordance with 10 CFR SO.SSa(a)(3)(i) reasonable to conclude that a moderately less rigorous inspection schedule would be capable of detecting any CL indications before they became large enough to be a concern.

Basis for Use: Conclusion (Continued)

Ff\lP - Units 1 and 2 incurs very minimal risk with extending volumetric intervals from two lSI periods to 10-yrs inspection intervals because there have been no service experience with cracking found in RPV CL OM welds, crack growth rates in RPV CL OM welds are small, and likelihood of cracking or through wall leaks is very small in RPV CL OM welds; therefore, the use of this proposed alternative will provide an acceptable level of quality and safety.

For these reasons, it is requested that the NRC authorize this proposed alternative in accordance with 10 CFR SO.5Sa(a)(3)(i).

Duration of Proposed The 4th lSI Interval.

Alternative:

Precedents: There are no previous precedents to this alternative.

1. MRP-349, "PWR Reactor Coolant System Cold-Loop Dissimilar Metal Butt

References:

Weld Reexamination Interval Extension", Enclosure 2 Status: Under NRC Review E1-4

Joseph M. Farley Nuclear Plant - Units 1 & 2 pr()DC)SE!(j Inservice Inspection Alternative FNP-ISI-ALT-13 Version 1.0 Enclosu re 2*

Materials Reliability Program: PWR Reactor Coolant System Cold-Loop Dissimilar Metal Butt Weld Reexamination Interval Extension (MRP-349)

"Note that due for double-sided printing of this document, blank have been into final document. The blank pages in this document were not an inadvertent omission.

~~f211

":1-elECTRIC POWER RESEARCH INSTITUTE Materials Reliability Program: PWR Reactor Coolant System Cold-Loop Dissimilar Metal Butt Weld Reexamination Interval Extension (MRP-349)

A Basis for Revision to the Requirements of MRP- 139 and American Society of Mechanical Engineers Code Case N-770 for Large-Diameter Welds at Cold-Leg Temperatures 2012 TECHNICAL REPORT

Materials Reliability Program: PWR Reactor Coolant System Cold-Loop Dissimilar Metal Butt Weld Reexamination Interval Extension (MRP-349)

A Basis for Revision to the Requirements of MRP- 739 and American Society of Mechanical Engineers Code Case N-770 for Large-Diameter Welds at Cold-Leg Temperatures I This documen t does NOT meet the reqUiremen~ t of 10CFR50 Appendix B, 10CFRPart 2 1, ANSI N45. 2- 1977 and/or the inten t of IS0900 1 !1994).

EPRI Project Manager C. Harrington r=r==lIf211 RESEARCH 1.:.- mcr." pow" IN,'Hr rU TE 3420 Hillview Avenue Polo Allo, CA 94304-1 338 USA POBox 10412 Polo AIIO, CA 94303-081 3 USA 800 .313. 3774 650.855.2121 (Jsleqi@epru :-om 1025852 www.er.i CO," Final Report, August 2012

DISCLAIMER OF WARRANTIES AND LIMITATION OF LlABILIl'IES THIS DOCUMENT WAS PREPARED BY THE ORGANIZATlON(S) NAMED BELOW AS AN ACCOUNT OF WORK SPONSORED OR COSPONSORED BY THE ELECTRIC POWER RESEARCH INSTITUTE, INC. (EPRI).

NEITHER EPRI, ANY MEMBER OF EPRI, ANY COSPONSOR, THE ORGANIZATION(S) BELOW, NOR ANY PERSON ACTING ON BEHALF OF ANY OF THEM:

(A) MAKES ANY WARRANTY OR REPRESENTATION WHATSOEVER, EXPRESS OR IMPLIED, (I) WITH RESPECT TO THE USE OF ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT, INCLUDING MERCHANTABILITY AND FITNESS FOR A PARTICULAR PURPOSE, OR (II) THAT SUCH USE DOES NOT INFRINGE ON OR INTERFERE WITH PRIVATELY OWNED RIGHTS, INCLUDING ANY PARTY'S INTELLECTUAL PROPERTY, OR (III) THAT THIS DOCUMENT IS SUITABLE TO ANY PARTICULAR USER'S CIRCUMSTANCE; OR (B) ASSUMES RESPONSIBILITY FOR ANY DAMAGES OR OTHER LIABILITY WHATSOEVER (INCLUDING ANY CONSEQUENTIAL DAMAGES, EVEN IF EPRI OR ANY EPRI REPRESENTATIVE HAS BEEN ADVISED OF THE POSSIBILITY OF SUCH DAMAGES) RESULTING FROM YOUR SELECTION OR USE OF THIS DOCUMENT OR ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT.

REFERENCE HEREIN TO ANY SPECIFIC COMMERCIAL PRODUCT, PROCESS, OR SERVICE BY ITS TRADE NAME, TRADEMARK, MANUFACTURER, OR OTHERWISE, DOES NOT NECESSARILY CONSTITUTE OR IMPLY ITS ENDORSEMENT, RECOMMENDATION, OR FAVORING BY EPRI.

THE FOLLOWING ORGANIZATION, UNDER CONTRACT TO EPRI, PREPARED THIS REPORT:

Westinghouse Electric Company THE TECHNICAL CONTENTS OF THIS DOCUMENT WERE NOT PREPARED IN ACCORDANCE WITH THE EPRI NUCLEAR QUALITY ASSURANCE PROGRAM MANUAL THAT FULFILLS THE REQUIREMENTS OF 10 CFR 50, APPENDIX BAND 10 CFR PART 21, ANSI N45.2*1977 AND/OR THE INTENT OF ISO*9001 (1994).

USE OF THE CONTENTS OF THIS DOCUMENT IN NUCLEAR SAFETY OR NUCLEAR QUALITY APPLICATIONS REQUIRES ADDITIONAL ACTIONS BY USER PURSUANT TO THEIR INTERNAL PROCEDURES.

NOTE For further information about EPRI, call the EPRI Customer Assistance Center at 800.313 .3774 or e-mail askepri@epri .com .

Electric Power Research Institute, EPRI, and TOGETHER ... SHAPING THE FUTURE OF ELECTRICITY are registered service marks of the Electric Power Research Institute, Inc.

Copyright © 2012 Electric Power Research Institute, Inc. All rights reserved.

Acknowledgments Westinghouse 1000 Westinghouse Drive Township, 16066 Principal Investigators W.

N. Palm This SPOllsof.ea by publication is a rev,"""r,..,

documenl that should literature in the Materials Reliability Program: PWR Reactor Coolant System Cold-Loop Dissimilar Butt Weld Extension

{MRP-349j: A Basis for Revision /0 the Requirements of MRP- 139 and American of Mechanical Engineers Case N-770 LargeDiomeler al Cold-Leg Temperalures.

EPRI, Polo CA: 2012.

1025852.

<< iii )

Product Description metal revision to MRP-139 and Code dOlcurne:nts are included in

Background

Alloy 82/182 been to periodic reexamination of cold-leg Alloy 82/182 DM butt welds essentially six or seven years, respectively. population includes various connections, reactor coolant (RCP) inlet and outlet (SG) and reactor vessel the Babcock steam supply are generally inspected from outside diameter accessibility and personnel radiation exposure depending on design. Only a limited number have DM welds in the SG nozzles access.

location, operating leg RV l1V,foL.l'" may not warrant urgent action.

Obiective objective of this project is to rlP\,rPln,n extending large-diameter Alloy cold-leg butt weld exam reexamination least 10 years. Extending past a 10-year changes to Section XI, and wi thin the scope of this program.

Approach approach to developing technical was to use the that had the to IS Results required no new information or new analyses. Instead, was compUed and documented in one clear, concise technical basis document demonstrating that a interval an level of safety Alloy 82/182 butt weld locations at cold-leg Applications, Value, and Use reexamination for cold-leg DM butt welds will enable the cold-leg exams to that is consistent with the interval provide additional exams.

Keywords Alloy 600 Alloy 82/182 Butt welds PWSCC

<< vi>>

Abstract Both Materials Reliability Program- (MRP-) 139 and American Society of Mechanical Engineers (ASME) Code Case N-770 require periodic volumetric reexamination of cold-leg Alloy 82/182 dissimilar metal (DM) butt welds susceptible to primary water stress corrosion cracking (PWSCC) essentially every six or seven years, respectively. This population includes various branch connections, reactor coolant pump inlet and outlet nozzles, steam generator outlet nozzles, and the reactor vessel (RV) cold-leg nozzles. Inspection of the large-diameter weld locations on an interval that is inconsistent with the interval of 10 years required by ASME Section Xl for other welds has resulted in hardship for utilities. While a consideration of selecting the six- or seven-year interval was to encourage utilities to perform mitigation of the affected DM welds and eliminate further PWSCC concern, operating experience to date suggests that the susceptibility of the cold leg RV nozzles may not warrant urgent action.

This technical basis demonstrates that the reexamination interval can be extended to 10 years while maintaining an acceptable level of quality and safety. This technical basis primarily uses existing work that has been extensively reviewed and accepted within the industry.

Therefore, this technical basis is sui table for use as a justification for the revision to MRP-139 and ASME Code Case N-770.

-< vii ~

Table of Contents Section 1: Summary of Technical Basis ..................... 1-1 1.1 Approach .................................................................. 1-1 1.2 Results ...................................................................... 1-1 1.3 Suggested ................................................... 1 Section Introduction .**.***.**..*.****.....*****.****.*******.*.2-1 2.1 Background ..............................................................2-1 2.2 Hardships with Removing Core Barrel ............................................................................ .

2.3 Scope and Objectives ................................................ 2-4 Section 3: Current Inservice Inspection Requirements ****************************************** 3-1 3.1 ASME Xl ....................................................... .

3.2 MRP-139.................................................................. 3-1 3.3 ASME N-770 .......................................... 3-2 Section 4: Service Experience for Cold leg Alloy 82/1 82 Bu" Welds *.******..******.******.*.********4 . . 1 1 Overview ................................................................. .

4.2 Cold leg Butt Weld locations ......................................4-1 4.3 Summary of Service Experience ................................... 4-2 Reactor Inlet and Outlet Nozzles .................... ..

Steam Generator Primary ................................ 4-3 Other Piping Weld Locations .................................. ..

4.4 Observations and Conclusions .................................... 4-5 Section 5: Deterministic Analyses: Flaw Tolerance of Cold leg Weld Regions ************************5-1 1 Overview ..................................................................5- 1 5 Previous Analyses ..................................................... .

5.2.1 MRP*44 -Interim Alloy 600 Safety Assessments for US PWR Plants .................................... 5-1 5.2.2 MRP-113 Alloy 82/182 Pipe Butt Weld Safety Assessment for U.S. PWR Plant Designs ............. .

Initiation and Growth Rate Comparison ....................... .

Critical .................................. ..

<< ix >

Crack Growth Analysis .............................................. .

5.3 Recent Analyses ...................................................... ..

5.3.1 Flaw Evaluation of CE Design RCP Suction and Discharge DM Welds .............................. .

Inlet Nozzle Flaw Tolerance Evaluations ................................................................ 5-8 5.3.3 Steam Generator Nozzle Flaw Tolerance Evaluations ...... ~ .... _..... "................. ~ ............. ~...... ........ **

• 5.4 Conclusions for Deterministic Analyses ....................... 0 Sedion 6: Probabilistic Analyses .............................6-1 1 Overview .................................................................. 6- 1 6.2 Probabilistic Fracture Mechanics Approach MRP 116 ................................................................................ 6-1 6.3 Statistical Approach ................................................... 6-4 6.4 Conclusions for Probabilistic Analyses .......................... 6-7 Section 7: Conclusions *....*.....*.....*.......*.....*...*..*.*...* 7 . . 1 Section 8: Proposed Revisions to MRP-139 and Code N-770 .***....**...*.**...****....**...*.* 8-1 8.1 Proposed Revision to Code N-770 Cold Leg Locations ................................................................... 8-1 8.2 Proposed Revision to MRP-l for Cold leg locations ......................................................................... 8-2 Section 9: References ****************..*************.*****.*****.** 9-1

<x>

list of Figures Figure Relatianship of Core Barrel to Reactor Vessel and Inlet and Outlet ................................................ ..

Time from LeaiKa<:~e Maximum Acceptable Initial Circumferential Flaws, Accounting for PWSCC and without Residual ............................................................. .

Figure 5-3 Maximum Acceptable Initial Circumferential Flaws, Accounting for PWSCC and FCG, with Fabrication Residual Stresses and an Inner Surface Weld Repair ................................................................... .

Figure Circumferential Flaw PWSCC Growth at RV Inlet OM Welds ........................................... 5-9 6-1 All Available OM Weld Results (7% Through-wall)- 1 ............................................ .

Figure 6-2 All Available Large OM Weld Inspection Results (7% Through-wall) - Case 2 .............................................. 6-6 Figure All Available OM Weld 1Il::>IJeCIIUfI "",,,'un,,

(7% Through-wall) 3 .............................................. 6-6

-( xi)

List of Tables Table 4-1 Typicallarge-diameter Alloy 82/182 Cold leg Butt Weld Locations ........................................................ .

Table 4-2 Summary of CracKing in Reactor Vessel Outlet Nozzles ......................................................................... .

4-3 Summary of CracKing in Japanese nAr"rr.r Inlet Welds ....................... ..

Flaw Summary-Welds ...................................................................... .

Table Crack Growth Analysis of Part-Circumferential Through-Wall Flaws in Cold Leg Bult Welds:

Westinghouse and CE Design Plants: Based on MRP-21 Crack Growth Rates ......................................................... 5-4 Table CracK Growth Analysis of Part-Circumferential Through-Wall Flaws in Cold leg Butt Welds: Babcock &

Wilcox Design Plants: Based on MRP-21 Crack Growth 5-4 Results Advanced Element Growth Analyses Circumferential Flaws ................................... _.

6-1 of 40-Year Probabilities ...................... 6-2 Table Inservice Inspection Sensitivity Study ....................... .

6-3 Summary of Probability of Cracking Results Hot and Cold Leg Welds .................................................. 6-4

.( xiii>>

Section 1: Summary of Technical Basis 1.1 Approach

'lnr,rr.*,,,h to developing this was to deterministic and probabilistic analyses that had been np,*tnr*rn~*rI Reliability (MRP) to develop the for service experience has compiled and to future cracking various butt weld locations. Work being the Pressurized Water Group (PWROG) to pump (RCP)

Combustion Engineering with flaw (RV) inlet nozzles was 1.2 Results summarized in report show that the flaw tolerance is very good 14 NPS) cold butt The time for growth due to is close to 10 years even for time to grow a flaw size of providing a detectable leak to the critical flaw size is in excess of 10 years and can be 40 years.

Probabilistic analyses, based on pro!oatHlls statistical methods, have shown wall growth in at higher TP",P'

inspection interval to 10 years while '""-11",U11.

safety.

-< 1-1 )

1.3 Suggested Revisions It is suggested, based on the results discussed above, that the inspection intervals of 6 and 7 years required by MRP-139 and Code Case N-770, respectively, for uncracked and unmitigated large-diameter cold leg Alloy 82/182 butt welds be revised to 10 years, consistent with the interval specified by ASME Section XI

[3].

< 1-2 ~

Section 2: Introduction

2.1 Background

Section XI of the ASME Boiler and Pressure Vessel Code [3] specifies a 10-year interval for inservice inspection of pressure-retaining welds. MRP-139 [1] and ASME Code Case N-770 [2] both require a more proactive periodic volumetric re-examination of cold leg Alloy 82/182 dissimilar metal (DM) butt welds, essentially every six or seven years. This population includes various branch connections, RCP inlet and outlet nozzles, SG outlet nozzles, and the RV cold leg nozzles. The branch nozzles are typical of the Babcock & Wilcox (B&W) and Combustion Engineering (CE) designs, are generally inspected from the outside diameter (OD), and have varying accessibility and personnel radiation exposure issues depending on plant design and environmental conditions. Only a limited number of US plants (one) have DM welds in the SG nozzles that are directly exposed to the primary water environment and thus fall within the scope ofMRP-139 and Code Case N-770. However, it should be noted that there are a large number of SG nozzles wi th Alloy 182 as a portion of the weld, though not exposed to the primary water environment and these welds may be required to be inspected as a result of an NRC condition on Code Case N-770. These SG nozzle welds will also typically be examined from the OD with plant-specific access and radiation exposure implications.

The RV cold leg nozzles are typically inspected from the inside diameter (ID) which requires that the core barrel be removed for access (See Figure 2-1). This exam, under ASME Section XI inspection requirements, occurs once per interval (10 years typically) which coincides with the inspection of the RV shell welds, thus minimizing core barrel removal evolutions. Inspection of these nozzles on a six- or seven-year interval requires removal of the core barrel solely for the purpose of performing these nozzle inspections. Removal of the core barrel should be minimized for a variety of reasons. As with any heavy lift operation, there are inherent risks to the personnel involved in the lift activities. Experience has shown that there are also risks associated with equipment damage including damage to the lift rig, guide studs, or the lower internals and reactor vessel itself.

Damage to these items has the potential to put plant personnel in further adverse situations along with significantly increasing outage time and radiation exposure.

<< 2-1 >

Outlet Nozzle Inlet Nozzle Core Barrel Figure 2-/

Relationship of Core Barrel to Reactor Vessel and Inlet and Outlet Nozzles 2.2 Hardships Associated with Removing the Core Barrel The removal of the reactor vessel lower internals assembly (core barrel) is considered to be a critical lift due to the weight of the component, the tight clearances involved, and the radiation emitted by the assembly. For these reasons, only the personnel directly involved with the movement of the internals are typically allowed in containment during the evolution. (Although site-specific procedures will vary, most will have much in common with the following

<< 2-2 )

description.) Remote cameras are utilized to allow most of the involved the lift to be outside of refueling cavity area to minimize Most lower internals solely by 1S to sit on floor of not to raise his head cab area the crane to maintain radiation as low as reasonably achievable (ALARA).

Communications are via portable to lifting lower internals, a "dry run" is typically performed where the crane is attached to the rig and placed onto the guide studs in the reactor Temporary markings are then made to provide alignment references for the reactor vessel. These markings are by crane and the crew to align crane to vessel. The ng 1S moved to the storage location and a second set Following completion the "dry run," the studs and the lower internals are latched onto the until This is for 10 minutes.

Following the internals are lifted out of the reactor and moved onto their many plants, removing core barrel requires that it be raised above the

....n'"'.lll'. cavity water level during the reactor vessel to stand location. As can be expected, the radiation exposure are very and necessitate work to for evacuation from containment and installation of shielding polar crane operator(s).

Additionally some are configured such that core portion remains exposed above the refueling cavity water level storage, often installation of shielding walls. walls severely limit the ability to perform other outage cavity maintenance activities and involve significant time and their handling.

susceptible to r.n.. ,,.,,Mnn,,1 impediments:

• <"'-,.-",,..... during the pre-job

• Clear affected by several and

• C()mmllrucat:tOI1S h,pn!!TPP'n all personnel were not or Evolution (IPTE) implementation additional enhancements are also identified to minimize the risk lower events. include: remote crane

<< 2-3 )

use oflaserlphotogrammetxy alignments, providing a load cell readout in the crane cab, using load cells with alarms, providing OE, and covering these events in continuing training.

Inspection of the reactor vessel nozzle welds from the ID is done remotely.

While OD exams may be possible at some plants, additional accessibility issues resulting in personnel safety and ALARA concerns arise for those plants where OD access is only through the floor of the refueling cavity, significantly limiting any benefits to this alternative.

2.3 Scope and Obiectives The objective of this project is to develop a robust technical basis for extending the large-diameter (~ 14 NPS) Alloy 82/182 RV cold leg, Rep, and SG cold leg butt weld volumetric exam re-inspection interval to ten years. This will enable the RV cold leg exams to be performed on an interval that is consistent with the interval for the removal of the core internals.

-( 2*4 >

Section 3: Current Inservice Inspection Requirem nts Alloy 82/182 Boiler Category B-F and welds, were not to applied to all dissimilar metal or similar metal More requirements to Alloy were developed and published in and most recently in ASME Section XI Code N-770 A cold butt re-examination frequency requirements is provided in the following sections.

3.1 ASME Section XI ASME XI Table IWB-2S00-1, andlor examination 100% of dissimilar metal nozzle-to-safe-end welds (Examination Category B-F) and dissimilar metal welds Category B-J). Section requires that these examinations be performed on a interval. Prior to 2007 this in "Inspection Program B." IWB-2412, "

provided a set of requirements inspection on a plant life. However, Inspection Program A was not in United States and was removed from XI Edition. The requirements for inspection on a in IWB-2411.

It should be noted that most in the have informed inservice inspection program for programs the number selected for examination the examination method. some cases, the in examined included the elimination of examinations of Alloy 82/182 DM welds were, at not known to have an active mechanism. However, the program not the ."'\P('T,(1,n inspection interval for welds that are examined.

3.2 MRP-139 MRP-139, Revision 1, "Primary System Piping Butt Weld

.uu,,,,,,,,,,,",," [lJ was published in December 2008.

<: 3-1 )

of weldments for fabricated Alloy 82/182 weld L.alteg'OrH~S E and I are applicable to welds at cold-leg temperatures. Category E weldments are defined as "those not with resistant have not been given an 81 (stress improvement) treatment, are than or equal to 4" NPS or serve an (i.e., B&W nozzles), and are exposed to cold leg temperatures." Category I weldments are defined as that are not and cannot be volumetrically inspected ... and are exposed to temperatures equivalent to cold leg temperatures."

Resistant are those considered to not be to PW8CC. Table 6-1 of MRP-139 specifies that the examinations of be performed once every six The requirements ofMRP-139 have been identified as "mandatory" in NEI-03 08, "Guideline for the of Materials Issues." Based on this "mandatory" classification, guidelines been implemented by all plants In United 3.3 ASME Code Case N-770 ASME N-770, Revision 1, "Alternative Examination and Standards 1 PWR Piping and Welds Fabricated With UNS N06082 or UNS W86182 Weld examination once examination every second examination period not to six years to seven years was made to be more compatible Code which are typically one third of the periods can years in duration. the basis document for N 770 [4J cited that the time for unmitigated cold leg to through wall ranges from 20 to 40 years 1 to Code Case was approved by Ve:cernb(~r 2009 a Revision 2 is in Neither these revisions has for of the cold leg weld 2 fur msoe(:oc)n tt*PrlI1Pt,.-., is 2011 the ofN-770, States. Following the AHA~"n'..

provided in the rule. Upon of 10 CFR50.55a, the MRP-139-1

""'-U""-' [18].

-( 3-2 ).

Section 4: Service Experience for Cold Leg Alloy 82/1 82 Butt Ids 4.1 Overview Alloy 82/182 butt in PWR plants a surface inspection as required by Section XI of the ASME Code as well as visual for acid These required since inception of piping inspection requirements of Section around 1980. As Ul"I..U""CU in Section 3, a more volumetric examination has self-imposed in the U.S. since 2005 through the requirements ofMRP-139 82/182 DM welds. All welds have now been examined at least once employing examination methods qualified in with L'-'-"H.I..LJ XI, Appendix 8.

rrplpNIj-pn inspections have been performed at PWR plants worldwide.

l ....lU"'UC;! of in Alloy 821182 or Alloy head np"ptr<lt'I('>n tPrY",p'r<ln'rp~ or A summary reactor welds is provided in the following sections.

4.2 Cold Leg Butt Weld locations location 14 NPS) Alloy leg temperatures in the Westinghouse, Combustion ,J"JUl'.HJ,\-'-'J and Wilcox plant are discussed in 3 and are summarized in 4-1.

-( 4-1 >>

Table 4-1 Typical Large-diameter Alloy 82/182 Cold Leg Butt Weld Locations Westinghouse Plants 1

,

• Steam Generator Outlet Nozzles 2 550-560 27.5 3 3

i . Reactor Vessel Inlet l'lozzles Combustion Engineering Plants

• Reactor Coolant Pump Inlet Nozzles4 30 4 549-560

• Reactor Coolant Pump Outlet 30 4 Nozzles4 i Babcock and Wilcox Plants 28 4

• Reactor Coolant Pump Inlet Nozzles 28 4

• Reactor Coolant Pump Outlet Nozzles 557 14 2

• Reactor Vessel Core Flood Nozzles 14 2

• Core Flood Tank Nozzle

1. Data is for a Westinghouse 3-loop plant. Number of typical locations is dependent on number of loops .

2. One Westinghouse plant has Alloy 82/182 butt welds between the reactor coolant piping and steam generator nozzles that are directly exposed to the reactor coolant.

3 . There are no Alloy 82/182 RPV nozzle welds in Westinghouse 2-loop plants and some early Westinghouse 3-loop and 4-loop plants.

4. Some CE plants do not have Alloy 82/182 RCP suction and discharge nozzle welds.

4.3 Summary of Service Experience All dissimilar metal (DM) welds in pipes 4" NPS and greater, including those containing Alloy 82/182, in categories B-F and B-1, have been subject to volumetric examination every 10 years, following the requirements of ASME Section XI. In some cases, these examinations were eliminated as part of a risk informed lSI program while in other cases they were supplemented by visual inspections for boric acid leakage. A summary of service experience [6] for Alloy 82/182 butt welds is provided in the following sections. Though there have been numerous incidents ofPWSCC identified in the pressurizer nozzle welds, this service experience is not included since these events have occurred at temperatures significantly higher (-653°F) than typical cold leg temperatures.

Reactor Vessel Inlet and Outlet Nozzles The only known incidents ofPWSCC in the reactor vessel inlet and outlet nozzles have occurred in the Alloy 82/182 nozzle-to-safe-end weld region of the outlet nozzle. These nozzles typically operate at 608 °F - 621°F. The first

incidents occurred in the outlet nozzles of Ringhals 3 and 4, and Virgil C.

Summer in the year 2000. Since that time, over 100 automated UT examinations of these welds in operating plants in the U.S. and internationally have been completed, typically coincident with the inspection of the reactor vessel shell welds. No additional surface indications were found until 2008, when indications were identified in the outlet nozzles of two different reactor vessels. The first was at OHI-3 in Japan. This indication was detected prior to the application of water-jet peening to mitigate PWSCC. The indication was measured by UT as being 10 mm in length and 5 mm in depth. When the indication was actually removed by progressive grinding, it was measured to have a length of 13.5mm and a depth of 20.3 mm. The cavity has been left in place. The second indication was detected at Salem Unit 1 as a result ofUT inspection, prior to the application of the mechanical stress improvement process (MSIP). This indication was determined to have a depth of -15 mm. Finally, in 2009, an indication was found in the Seabrook reactor vessel outlet nozzle. This indication was axially oriented with a depth of -15.6mm and a length of -24.4mm. A summary of the incidents ofPWSCC found in the reactor vessel outlet nozzles is provided in Table 4-2 Table 4-2 Summary of Cracking in Reactor Vessel Outlet Nozzles 621 Seabrook 621 OHI3 617 14.0 Ringhals 3 613

• 12.8

~--~-~--

Ringhals 4 613 12 .3

--~~--,.j Salem 1 608 19.7 *

1. Effective Full Power Years of Operation at the time the indication was found.

Steam Generator Primary Nozzles Cracking in the steam generator nozzles has only been observed in the Alloy 82/132 inlet nozzle-to-safe-end weld region of steam generators in Japan. For plants in the U.S. that have stainless steel reactor coolant system main loop piping, steam generators were originally fabricated with stainless steel nozzle-to safe-end welds. Many plants have replaced their steam generators and in doing so have installed steam generators with either stainless steel welds, or welds fabricated with Alloys 52 and 152, which are considered to be relatively unsusceptible to PWSCC. One plant in the U.S. does have Alloy 82/182 welds in the steam generator nozzle-to-safe-end welds. This plant recently (spring 2012) mitigated the inlet (hot leg) welds and identified cracking, presumed to be PWSCC, in one nozzle. In some cases Alloy 82/182 welds were used with a layer of Alloy 52/152 to seal the Alloy 82/182 material from the primary coolant water. As a result of a condition imposed by the NRC on Code Case N-770 in

..( 4-3 )

10 CFR 50.55a, owners are required to have NRC approval before such welds can be considered as mitigated and not in the scope of Code Case N-770.

In Japan, most steam generators were originally fabricated with Alloy 132 nozzle to-safe-end welds. Alloy 132 is similar to Alloy 182 and is equally susceptible to PWSCc. Therefore, the Japanese PWRs with susceptible welds are implementing peening as mitigation for these welds. In preparation for peening, the inside surface of the welds must be inspected. While these inspections (and subsequent peening) had been successfully applied at five plants, during the inspections ofMihama 2 and Tsuruga 2 in the fall of 2007, indications were detected. In November of 2007, NISA, the Japanese regulatory authority, issued a guideline for each susceptible unit to inspect the nozzle-to-safe-end weld region at their earliest convenience. As a result, five additional plants have detected cracking in this region. All indications have been detected in the inlet nozzle-to-safe-end weld region, which is the hottest location, typically operating at 608°F - 621°F. Note that no indications have been found in the outlet nozzles, which operate at cold leg temperatures. A summary of this experience is provided in Table 4-3.

Table 4-3 Summary of Cracking in Japanese S/eam Generator Inlet Nozzle-to-SaFe-End Welds Number of Indications, Max. L, Max D Plant Date A. Loop 8 Loop C loo~

13 indications Mihama Unit 2 September 500MWe 2007 l= 17mm o indications N/A D= 13mm i

Tsuruga Unit 2 ! - b 1 N ovem er 1 indicatio-n-s-r--5 indications i 23 ir1dications I l=N/A l=21 mm l=14mm '

1110MWe : I Ir I 2007 D=N/A D= 12mm i

, D= 13mm

, Takahama December 3 indications 2 indications 4 indications ~

Unit 2 l=7mm l=7mm l= 11 mm 2007 780 MWe .,..~~-

D=N/A D=6mm D=8mm 3 indications I,' Genkai Unit 1 i l=5mm o indications N/A i 529 MWe January 2008 D=N/A Takahama 7 indications 16 indications 9 indications February Unit 3 l=28mm l=38mm l= 14mm 2008 870MWe D=9mm D= 15mm D=9mm I Tomari Unit 2 I' 579 MWe ,

Takahama Unit 4 April 2008 October 2008 r

I indications L= 13mm D=7mm 7 indications l= 14mm 10 indications l= 1Omm D=5mm 8 indications l=30mm N/A L=33mm J 21 indication7!

870MWe D=12mm D=13mm D=16mm

--~~-

D = Depth,_l=~~n~t~, ~/ A = !'Jot ~'p~cab~ ~

-( 4-4 >

Other Piping Weld Locations In (CE) and Babcock Wilcox (B&W) 182 or Alloy 82 butt welds used to join steel (instrumentation drain to the main loop piping, which is steel. There have numerous incidents of cracking in these locations. found predominantly in the high with very few incidents of cracking in colder locations [7].

However, few in the colder have in welds with diameters of less than 14 NPS which do not have the flaw tolerance of the welds that are 14 NPS as will be in report.

these welds have not been included in the proposed change in inspection UJ.o."u.,o<,\#U in this 4.4 Observations and Conclusions following conclusions can be drawn from the above experience:

.. It can be concluded known of cracking in bore Alloy 821182 piping welds occurred in locations operating at hot or

.. No safety or structural integrity concern has resulted from cold leg butt weld PWSCC to

( 4-5 .>

Section 5: Deterministic Analyses: Flaw lerance of Cold Leg Id Regions 5.1 Overview response to u.a'~""!l'" incidents in 4, a number analyses were to assess stability of piping with PWSCC determine the extent through-wall analyses were UV'_UU1'-'UL\.-U in the reports below and as basis for the and evaluation guidelines identified in MRP-139.

5.2 Previous Analyses 5.2. I MR.P Interim Alloy 600 Safety Assessments for US PWR. Plants MRP-44 [8], published in April 2001, provided interim safety assessments most susceptible Alloy 82/182 weld locations in US PWR locations included the reactor vessel to hot Westinghouse line welds for engineering welds for the Babcock and Wilcox oriented flaws and through-wall over a relatively around the provided II If cracks develop in welds, they are expected to be predominantly axial.

II Axial cracks in by low-alloy steel or steel at either the weld are limited to the width of the weld. The critical flaw size for rupture is times than width of the II Through-wall circumferential leaks can be detect(~d structural ... _.,..,..._.

<< 5-1 )

NRC ,.,p,*t",*YY'I"ti a review of report and concluded that it a basis for continued while additional analyses and inspections are performed. It should conclusions were reached for the evaluation oflocations hot 5.2.2 MRP-I J3 - Alloy 82/J 82 Pipe Buff Weld Safety Assessment for U.S. PWR Plant Designs provides the final safety assessment PWSCC of Alloy welds in PWR plant primary It is a continuation of work documented in MRP-44. The conclusions report are supported by

"""'"'.,,'" and contained in a of supporting MRP reports

'-H-''I<. but not limited to MRP-109 MRP-112 [10], and 16 [11].

provides to assess butt weld PWSCC in Westinghouse and Combustion Engineering plants provides the analyses Babcock and Wilcox MRP-116 provides the results fracture mechanics np,*tro,'m"ri to assess the probability the probability to crack growth, and change in core damage frequency (CDF) from PWSCC of Alloy 821182 butt Though the results ofMRP-116 are summarized in MRP-I13, results ofMRP-116 will be discussed in Section 6.0 In;t;at;on and Growth Rate Comparison are several factors that influence the initiation and growth of cracks in Alloy 82/182 weld The most significant include the susceptible material, the tensile stress, and the environment, As discussed in MRP-113, the general is that, PWSCC susceptibility with applied tensile stress, the time to LU","CH'" of the operating Locations operate at such as the cracking sooner than locations that operate at lower temperatures, such as in RCS legs. For typical PWR plant pressurizer (653°F), hot (600°F), and cold (550°F) and a activation of 50 kcallmole initiation, on time to initiation ofPWSCC for hot leg and leg locations relative to pressurizer locations are 7.7 and respectively. If predictions are based on crack growth rate data, the activation can be as 31 kcallmole the multipliers on time are and respectively. In words, under conditions, cracks in cold leg locations 63.7 times as to initiate and grow at a rate 13.1 times slower than locations.

Critical Crack Size Assessment Westinghouse (MRP-109) and AREVA have n;>.-~..,.,.nPrl <.1"<.1,,,,,,,,,,

sizes a range ofAlloy 82/182 nuclear steam supply system (NSSS) ""n,.,I1"*",,

on ASME

calculated for both the circumferential and axial orien tations, recognizing that experience indicates that axial flaws are limited to grow by PWSCC to the width of the weld. These analyses were performed to determine the most limiting conditions for the plants in the domestic fleet. The results of the limiting crack sizes are shown in Table 6-1 of MRP-113 while the results pertaining to cold leg locations are summarized in Table 5-1 of this report. The data in Table 5-1 shows that Alloy 82/182 cold leg butt welds in domestic PWR plants can tolerate axial and circumferential flaws of a significant size while maintaining structural integrity.

Table 5-7 Critical Flaw Size Assessment Summary - Cold Leg Welds 115 .66 RPV Inlet 7.7 25.9 13~_J____ ~~

RPV Core Flood B&W A 22.3 194 I .75 SG Outlet W D 8.8 30.0 155 .77 RCP CE J 9.4 38.2 115 .62 Suction I RCP 9.4 38.2 104 .56 I Discharge

1. These critical axial flaw lengths are much greater that the width of the Alloy 82/182 butt welds.

Crack Growth Analysis No calculations were performed for the growth rate of axial flaws since the analysis results demonstrated that the maximum lengths of through-wall axial cracks, which are limited to the width of the Alloy 82/182 weld, are significantly less than the calculated critical crack sizes. Westinghouse and AREVA performed crack growth analyses for circumferentially oriented cracks. These analyses were originally performed using the weld crack growth rate model in MRP-21 [12]. Mter the final crack growth model was published in MRP-115

[13], check calculations were then performed for the limiting cases. These cases confirmed that the results obtained from the analyses using the MRP-21 crack growth rate model were conservative. While there are differences in the approaches taken by Westinghouse and AREVA, the results from both approaches show that the flaw tolerance in cold leg weld locations is very high.

< 5-3 :>

The results of the analyses performed by Westinghouse and AREVA for the Alloy 82/182 butt weld cold leg locations are summarized in Tables 5-2 and 5-3 respectively.

Table 5-2 Crack Growth Analysis of Part-Circumferential Through-Wolf Flaws in Cold Leg Butt Welds: Westinghouse and CE Design Plants: Based on MRP-21 Crack Growth Rates r RPV lolet I SG Outlet I w D > 40 > 40 > 40

, RCP i J 27.0 > 40 > 40 > 40 Suction RCP Discharge J 19.7 > 40 > 40 38.5

1. Aspect ratio defined as: Flaw length:Flaw depth.

2. Through-wall crack producing either 1 GPM or 10 GPM leak.3.

Table 5-3 Crack Growth Analysis of Part-Circumferential Through-Wa" Flaws in Cold Leg Butt Welds: Babcock & Wilcox Design Plants: Based on MRP-21 Crock Growth Rates As can be seen by the results in Tables 5-2 and 5-3, the times for growth of postulated flaws to limiting size in cold leg butt weld locations are very long. In most cases the results show that more than 40 years is required to reach 75-100%

through-wall. These long times result from the crack tip stress intensity factor dropping below the MRP-21 threshold of 9 MPa."lm for PWSCC crack growth.

The only growth predicted under these conditions would be by fatigue.

Westinghouse and AREVA performed additional crack growth assessments using the MRP-115 crack growth rate model without the stress intensity factor threshold. The data show that the times for cracks to grow through-wall are reduced. However, the times for cracks to grow from a 1 GPM or 10 GPM leak

-( 5-4 >

to length are increase results the that new crack growth rates are lower than original rates at higher K levels.

cases, the were performed using same initial flaw sizes as earlier analyses.

5.3 Recent Analyses 5.3. J Flaw Evaluation of CE Design RCP Suction and Discharge Nozzle DM Welds reactor coolant inspection coverage '-U,CLU'-,H!','""

required inspection coverage are in WCAP-17128-NP, Revision 1 [14]. ",",,,,au,,,,,, were divided into 3

1) 2) ASME Flaw and 3) Advanced Tolerance.

Defense-in-Depth To a measure region, calculations were oertorme:C1 tiptprt1'm and the time through-wall 111 ....lU'~"'U m to growth Considering that the by plant technical specifications for most plants in the U.S. is better than GPM, even if a flaw grew through-wall, the time in it can be e:Jq)ected to grow to a critical length is 10 years. Since the actual rate sensitivity is closer to 0.1 GPM, the time to grow to a critical length 15 years.

-( 5-5 >>

Tim 10 Reach Crl~aI Lengthvs. lritlal Crack Leakage Rale 20~--------~--------~----------r-------~==~==~==~

18

__ 1 -+-

Mlrirrum NoP Load Maximum NoP Load 16

..~ 12 10 T

.5 1

4 I

o ~ ________ ~ __________- L_ _ _ _ _ _ _ _ _ _ ~ __________ ~ _ _ _ _ _ _ _ __

o Leak Rille (gpm)

Figure 5-1 Time from Leakage to Critical Circumferential Flaw Length (No Residual Stress Case) for a Through-wall Flaw ASME Section XI Flaw Tolerance A series of flaw tolerance calculations were carried out in WCAP-17128-NP, Revision 1 [14] to determine the time required for a postulated surface flaw to reach the ASME Section Xl allowable flaw size. Both fatigue crack growth and stress corrosion cracking were considered, and the results were presented in terms of the allowable service time for a range of flaw sizes and shapes. The calculations determined the range of flaws which are acceptable for service periods from two to four years. These calculations include the required Section Xl flaw evaluation margins and were presented for both axial and circumferentially oriented flaws.

Residual stresses were calculated using finite element analysis techniques [5]

assuming cases of no weld repairs and weld repairs of different through-wall depths up to 50% from the ID. The results for the circumferential flaws show that very large flaws can be tolerated in this region as the residual stress effects were found to retard flaw growth for circumferential flaws (i.e., the results for the cases without weld repairs are more limiting). While the results for the axial flaws do not exhibit as much tolerance as for circumferential flaws, the limited length of the flaw causes the aspect ratios to also be limited. Though not included in WCAP-17128-NP, additional analyses consistent with those described above were performed for circumferential flaws for a service period of 10 years. The results of these evaluations, with and without residual stresses due to weld repairs, are shown in Figures 5-2 and 5-3, respectively. These results show that flaws with an aspect ratio as large as 10 and a through-wall depth of 20% will be acceptable for at least 10 years.

< 5-6 )

0.9

~ ~J 0.8

~

~0.7 _ _ _I a::

~ 0.6 ___,.-----~- . :=:=:-==-- I c:

"".S!

F= 0.5

~~

13-...-..

.----'...a-----

---

==. '.- ----

a.-.- ------ -.-

_____..--- --~ j!

i5. .....-.~ i

~ 0.4

~ ... _----- ..... - [I

<.> Time (months) to Reach 5 0 .3 n;

]02 0.1

--- ._ ~-.:: ASME Al lowable Crack Depth

......... 24

-<>- 36

~48

.. *..* 120 0

0 0.1 0.2 0.3 0.4 0.5 Crack Depth I Length Ratio . alt Figure 5-2 Maximum Acceptable Initial Circumferential Flaws, Accounting For PWSCC and FCG, without Residual Stresses i

0.9

~

O.S I

I

.~' 0.7

---

-.~=

!Z+/-3 _ L t

a::

~ 0.6

~;:::;:...'

,r

.... ~--

-- - -' ---- t c:

""<.> ,/

~O. 5 i5.

~0 .4 Tlme(monlhs)loReach I

""<.> 0 3 5

~

. IASME Allowable Crack Depth .

.s" 0.2

-+- 24

- ! - 36 1 0.1 --a 48 1 - -** 120 0

0 0.1 0.2 0.3 0.4 0.5 Crack Depth I Length Ratio, alt Figure 5*3 Maximum Acceptable Initial Circumferential Flaws, Accounting for PWSCC and FCG, with Fabrication Residual Stresses and an Inner Surface Weld Repair

-( 5-7 ~

Advanced Flaw Tolerance Analysis The ASME flaw tolerance work was supplemented with advanced finite element analyses, wherein the postulated flaw was allowed to grow due to PWSCC in a natural shape, dictated by the stresses present. The results of these analyses are shown in Table 5-4 and are based on a postulated surface flaw in the region which cannot be inspected, with length equal to 14% of the circumference. The depth of the flaw was varied from 20% to 30% of the wall, to bracket the range of uninspectable materials. These depths were chosen based on very conservative aspect ratios of 0.04 and 0.03, respectively. These are significantly larger than the aspect ratio of 0.1667 typically observed in service, and it is highly likely that any flaws deeper than this would have tails which would be detected in the inspected region. Results show that the postulated flaw will remain within the ASME Code acceptable depth for 7.5 to over 11 years, depending on its initial depth, and requires between 9.3 and 13 years to reach a through-wall condition. These results do not account for the beneficial impact of the stainless steel field weld that is made to join the cold leg or crossover leg safe-end to the stainless steel RCP casing. This weld induces a region of compressive stress in the mid wall region of the pipe, which would further retard the crack growth.

Table 5-4 Results of Advanced Finite Element Crack Growth Analyses for Circumferential Flaws 0.20 9.6 years 11 .1 years 0.30 0.14 7.44 years 9.34 years 0 .30 0 .23 6.45 years 7 .85 years Summary This work documented in WCAP-17128-NP, Revision 1 [14J has demonstrated that the pump safe-end to nozzle weld regions have Significant margins, and therefore do not require the accelerated inspection frequency specified in MRP 139 [1] and Code Case N-770 [2]. The three approaches used to support this conclusion have been consistent in their findings.

5.3.2 Reactor Vessel Inlet Nozzle Flaw Tolerance Evaluations Westinghouse has performed a generic flaw tolerance evaluation to determine the maximum flaw sizes in the reactor vessel inlet dissimilar metal welds that would support continued operation for a period of 10 years. This evaluation was performed consistent with the ASME Section XI flaw tolerance evaluations performed for the RCP nozzles as discussed in Section 5.3.1. Along with the normal operating steady state piping loads, the impact of welding residual

~ 5-8 >

stresses under different safe end lengths and the various extent of inside surface weld repairs during the initial weld fabrication process were considered in the evaluation. These residual stresses were also calculated using finite element analysis techniques that are consistent with recent industry guidance [15]. A parametric study was performed to evaluate the residual stresses for the different weld and safe-end configurations present in the Westinghouse fleet. Based on a comparison of the various residual stress distributions from the parametric study, it was concluded that a long (Length> 4.5") safe end with either a 25% or 50%

inside surface weld repair would produce limiting PWSCC crack growth results.

A high and a low cold leg operating temperature were also considered in the evaluation to represent the range of operating temperatures in the fleet [16].

Based on the circumferential crack growth results shown in Figure 5-4, even for the most conservative case (high temperature with a 25% weld repair) a flaw with a depth of 15% of the wall thickness would not grow to the maximum allowable ASME flaw size in less than 10 years of continued operation. I t should be noted that the results presented in Figure 5-4 are not representative of a single plant.

These results are based on the limiting thickness in the Westinghouse PWR fleet combined with the limiting piping loads from another plant in the Westinghouse PWR fleet and therefore, these results are conservative.

Cold Leg Cire Flaws 0.9 ..... M _* ** _ _ _ **** ~ ** _ .. _ . ........- - * *** ~- ** - * * ** - -**- . -" - ' - - -" - -- " ' - -- " - ' - **** * * ** _ ******* _. _ *** _. __. - *****--*--*****-**-- - -*Sliort*S£*7"5'CJi'FfipaT;*-*-- - *-*****-***._ _ ..........__..

~~"Q~g~T~~~a~f AR= lQ,HlghTamp 0 .8

/

,/

0 .7 Ena-of-EvaluaUon PeriOd FlQiw SIZe =0 57 c:

~ 0.6

c l

+-______-r"--_ _ _ _ __

Long SE 25% Rej)Sllr ia. 0 .5 -=.",..,~=--------_ AR= 10, Law Temp 0 0.4

~

u:

0.3 0.2 0.1 0

0 5 10 15 20 25 30 35 40 45 Tim. (Ye... )

_____ High Temp Sh ort SE 25% repair AR = 10 - - High TempLongSE 25% rapair AR = 10 HlghTemp = 565 F

.... - .._. Low Temp Long SE 25% repair AR = 10 Low Temp = 535 F

-*-LowTemp ShortSE 25% repair AR =10 Note: AR = Aspect Ratio, SE = Safe-End Figure 5-4 Circumferential Flaw PWSCC Crack Growth at the RV Inlet nozzle DM Welds

-( 5-9 >

5.3.3 Steam Generator Nozzle Flaw Tolerance Evaluations As indicated in Section 4.3, cracking has recently been identified in one steam generator inlet nozzle in the U.S. and but flaw tolerance evaluations have not been performed. Results for the steam generator outlet nozzles could be expected to be similar to those reported above for the reactor coolant pump and reactor vessel cold leg locations.

5.4 Conclusions for Deterministic Analyses All of the flaw tolerance analyses performed to date have shown that the critical crack sizes in large-diameter butt welds operating at cold leg temperatures are very large. Assuming that a flaw initiates, the time required to grow to through wall is in excess of20 years in most cases analyzed. The time to grow from a through-wall leak to a crack equal to the critical crack size can be in excess of 40 years.

More recent analyses have been performed for the RV nozzles using through wall residual stress distributions that were developed based on the most recent guidance. These analyses have shown that the flaw tolerance of these locations is high and postulated circumferential flaws will not reach the maximum ASME allowable depth in less than 10 years. Supplemental advanced finite element analyses performed for the CE RCP suction and discharge nozzles shows that even if a large flaw is assumed to exist, the time to grow through-wall is a minimum of approximately 8 years. Furthermore, this flaw would be expected to take at least 10 years to grow to a critical length.

<< 5-10 )

Section 6: Probabilistic nalyses 6.1 Overview All of been to determine a can be performed using statistical methods. These two approaches are following sections.

6.2 Probabilistic Fracture Mechanics Approach - MRP-116 of the original to develop MRP-139 requirements, a probabilistic safety assessment was performed by Westinghouse for r1r>lmp,.ttr Westinghouse, Combustion Babcock & Wilcox (PFM) work is results are provided in MRP-116 [1 assessment was in 2004, it is most recent susceptible welds of different sizes and operating have been in of variables that effect PWSCC, the assessment still provides valuable insights into the likelihood weld due to PWSCC.

The probabilistic assessment on the deterministic work and probability that a flaw could grow through wall and could eventually lead to and a resultant increase in core damage documented in the report were intended to cover all the PWRs in the USA. The probabilistic safety assessment brings results, as well as complementary work to provide input on the of repairs and crack growth modeling.

Probabilistic fracture mechanics evaluations were performed to identified degradation of PWSCC and dissimilar metal butt butt welds in deterministic \OV""U"UVl RV inlet were not determined to limiting locations in deterministic Evaluations for of the considered the small

<: 6-1 )

small circumferential leak failure modes, and can be conservatively used to represent the results for cold leg locations. The results of the PFM evaluation for the circumferen tialleak probabilities, which represent a direct safety concern, are summarized in Table 6-1.

Table 6-1 Summary of 40-Year Leak Probabilities Decay Heat

~ B&W S.OOE-OS RV Outlet Nozzle W I, 2.00E-04 Safety/Relief CE/W

~-

9.S1 E*06 SDC CE 2.70E-OS

---

--~--.~

SG Inlet CE 3.3SE-06 Spray CE/W 1.2SE-04 Surge HL CE 3.3SE-06 CE/W 2.00E-04 Surge PZR B&W 2.00E-04 As shown in Table 6-1, the circumferential leak probabilities at 40 years are small. It must be noted that all of these probabilities are for cases evaluated at hot leg or pressurizer operating temperatures. Though not explicitly evaluated, based on the differences in crack initiation and growth times discussed in Section 5.2.2, the probabilities for locations at cold leg temperatures would be expected to be at least an order of magnitude less than those for the welds at hot leg temperatures, and higher.

As part of the MRP-116 probabilistic fracture mechanics evaluations, a sensitivity study was performed to determine the effects ofISI accuracy and frequency. This sensitivity study was performed for a weld that was considered to be representative of the welds included in the study. The results of the study are shown in Table 6-2. Though the weld considered in this study was not a cold leg weld, the results of the study would be expected to envelope the results for cold leg weld locations.

~ 6-2 >

Table 6-2 Inservice Inspection Sensitivity Study I No ISI 2

---

~--------~

10 Year ISI 2 1 Year ISI 2 1 Yr lSI and Improved Quality3 of.

Inspection r- ---- . .

1. Residual stress input unchanged

2. Standard inspection quality for 50% detection of a flaw 25% through the wall

3. Standard inspection quality for 50% detection of a flaw increased to detect a flaw 10% through the wall

4. For conditional core damage probability (CCDP) = 3.0E-03 Based on the results of the probabilistic fracture mechanics analyses, it was concluded in MRP-116 that:

• Changes in inspection frequency or improvements in capability or accuracy have only a small benefit for the locations with the highest leak probabilities.

• Risk results do not justify shortening the current 10-year ASME Code Section XI inspection interval, as long as all Alloy 182/82 locations are included.

NRC Regulatory Guide 1.174 [17] provides An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Curren t Licensing Basis. Regulatory Guide 1.174 defines an acceptably small change-in-risk as one that meets the following criteria:

• Change in Core Damage Frequency (CDF) < 1 x 10-6 per reactor year

• Change in Large Early Release Frequency (LERF) < 1 x 10-7 per reactor year Based on the results shown in Table 6-2, the change in risk (CDF) in moving from a hypothetical 1 year lSI interval to the ASME Section XI 10 year lSI interval would be 1.69 x 10-9 per reactor year. This change is more than 2 orders of magnitude below the regulatory criteria for an acceptably small increase in risk.

It is reasonable to conclude that the increase in risk in moving from a 6 or 7 year interval as required by MRP-139 or Code Case N-770, respectively, to a 10 year interval would be even less than 1.69 x 10-9 per reactor year and further acceptable per Regulatory Guide 1.174. Furthermore, it would also be expected that the change would be less for cold leg weld locations. In other words, the shorter intervals specified in MRP-139 and Code Case N-770 are not needed for the cold leg locations to satisfy risk objectives.

< 6-3 >

6.3 Statistical Approach A probabilistic analysis was performed in WCAP-17128-NP, Revision 1 to assess the susceptibility of cold leg welds to PWSCc. The analysis considered available industry experience data for the locations of Alloy 82/182 DM welds.

More specifically, the data analyzed included Alloy 82/182 DM welds that were nominally 28 inches in diameter or larger at the:

1. Reactor vessel inlet and outlet nozzles,

2. Steam generator inlet and outlet nozzles, and

3. Reactor coolant pump suction and discharge nozzles.

In addition to the service experience data for the above large nozzles, service experience for the pressurizer surge nozzle was also analyzed in one case .

The collected service experience data was fit to a Weibull distribution which was then used to calculate the probability of cracking as a function of EFPY. This was done for three different temperatures with the intent of covering the range of temperatures on the cold nozzle DM weld locations (548°F to 556°F), as well as a representative hot nozzle DM weld location (615°F). Three different cases were evaluated based on the data to which the Weibull distribution was fit. Case 1 is based on all the available inspection results, for reactor vessel nozzles, steam generator nozzles, pump nozzles, and pressurizer surge nozzles. Case 2 includes all the nozzles except the pressurizer nozzles, and Case 3 includes only the reactor vessel and RCP nozzles. The results of these cases at the three temperatures are shown in Table 6-3. The cumulative probability of cracking with respect to effective full power years of operation was also determined for each of the three cases and is shown in Figures 6-1 to 6-3.

Table 6-3 Summary of Probability of Cracking Results for Hot and Cold Leg Welds AtE'PY Cas81 Cas82 Case 3 Temperature 548°F

- '--:--r 0.25% 0.00% i 0.01%

40 0.57% 0.03% 0.05%

60 0 .93%

1 0.12%

1 0 .15%

0.02%

40 0.10% 0.13%

---r 60 I 0.35% 0.35%

Temperature 615°F i

i 20 6.98% 20.92% 9.84%

40 15.32% 86.63% 44.34%

I 60 23.71 % 1 99.92% 80.10%

-( 6-4 :>

The results in Table 6-3 show that there is no discern able difference between the cases at the cold leg temperatures. Furthermore, the predicted probability of cracking for DM welds operating at cold leg temperatures is extremely low, even at 60 effective full power years (EFPY). The results of the Weibull curve fitting for the three cases indicate that even though DM welds have had many flaws at hot temperature locations, none have been found at cold temperature butt weld locations, and this gives a very low probability of flaws existing in cold temperature locations. Results in Table 6-3 show that the highest probability of an indication at cold leg temperatures was only 1.42%, at 60 EFPY (Case 1 at 556°F). A 60 EFPY value is beyond a plant's licensed life, even with a 20-year life extension.

The cumulative probability of cracking with respect to effective full power years of operation was also determined for each of the three cases and is shown in Figures 6-1 to 6-3.

All Available Large OM Weld Inspection Results (@7% tw) 100%

90%

-1 -" --,---- - - ,- -

Welbull Parameters 80'4 Shape: 1.2 Scale: 324 EDY

~ 70'!.

sco

.Q e iI:

0.. co 60%

,,0::

~~

.;~

e .... so'"

> co -615 0'0

~

40% - -556 I

.Q

.; - 548

~ 30'4 20'!.

10'4 L------ ~ -

0'4 10 20 30 40 50 60 Effective Full Power Years (EFPy)

Figure 6-1 All Available Large OM Weld Inspection Results (7% Through-wall) - Case 1

<< 6-5 ~

RV, RCP, and SG Large DM Weld Inspection Results (@7% tw) 100%

90%

-T , - - ---,

-~

Wei bull Parameters Shape: 3.1 /

_L 80%

Scale: 58 EDY

~

c.. 70%

.Q ir. .

0

~

.. '" 80%

I

/

~!

-5~ 50% J 00 E'"

40%

/ -615

-556 - '

.Q

'CD 3: 30% L 548 20%

/

/

-V 10~.

0%

10 20 30 50 60 Effective Full Power Years (EFPy)

Figure 6-2 All Available Large DM Weld Inspection Results (7% Through-wall) - Case 2 RVand RCP Large DM Weld Inspection Results (@7% tw) 100% _ .'-' ,- ..-.. .. - . ~

90%

Weibull Parameters 80% Shape: 2.5

~

c 2l 70%

Scale: 90 EDY V

Z e

..... ~ L

.. '" 60%

~!

!!!';I!.

50%

/'

80 E ..

./

./ -615 40% -556

.Q

'CD 3: 30-1.

/'

/' - 548 20'"

/

10'10

/

O'Y.

~

10 20 30 40 50 60 Effective Full Power Years (EFPy)

Figure 6-3 All Available Large DM Weld Inspection Results (7% Through-wall) - Case 3

-< 6-6 >

6.4 Conclusions for Probabilistic Analyses the probability of failure Alloy u,,-,_c,,",,", mechanics statistical methods.

likelihood of cracking or through-wall Furthermore, sensitivity that even

<lhc,nll1tp HIUJ.'_'<JA of the likelihood service in large-bore buttwe1ds all indications ifhot inspections any cold concern.

-< 6-7 )

Section 7: Conclusions While there has been a large amount with primary water stress corrosion cracking of Alloy 82/182 buttwelds, this has been limited to welds leg or higher (60goF with few exceptions. have been no incidents of cracking in butt welds operating at cold leg temperatures ( < ) that can be attributed to PWSCc. The MRP-139 and Code requirements for more were as a proactive measure.

accumulation of more positive service indicates that while this hot weld locations a hardship to and m a y a n to the complications with removal of the have been numerous evaluations performed of likelihood of through-wall in cold leg Alloy 821182 The MRP-l39 showed that the probabili ty residual stress distributions, ViAj","""'" conclusions flaw tolerance is high.

more recent analyses have shown that even large circumferential flaws, with a high likelihood of being during inservice will not grow to the maximum depth allowed by ASME Section XI in 10 years. analyses have been performed based on the that a flaw which as shown more recent probabilistic based on service data is unlikely at the present time.

It is concluded, that an interval of 10 years re-examination diameter cold Alloy 82/182 will provide a more than adequate level of safety and Furthermore, this interval will hardship on utilities and minimize overall plant risk associated with movement of the reactor core barrel. Proposed revisions to MRP-139 and Code N-770 to incorporate a lO-year re-examination interval for cold leg butt welds are shown in Section 8.

<. 7*1 )

Section 8: Proposed Revisions to MRP-139 and Code Case N-770 The inspection of Alloy 182/82 DM welds since 2005 have been performed to the requirements of report MRP-139, Revision 1 (Reference 1). These inspection requirements have now been replaced by those of Code Case N-770 (Reference 2).

The proposed revisions to Code Case N-770 and MRP-139 to require an inspection interval of 10 years are discussed in the following sections.

8.1 Proposed Revision to Code Case N-770 for Cold Leg Locations It is proposed that Inspection Item B ofTable 1 of Code Case N-770 be revised as follows:

Existing Table 1 Requirements for Inspection Item B:

Unmitigated Once per butt weld at 1 I interval Cold leg Every second operating Weld Surface I Visual (2), (3) -3140 Not B

I i

temperature (-

2410) 2: 525°F Fig . 1  ! Volumetric (4) -3130 inspection period not to I Permissible I

exceed 7 yr I ~2:0~~C(~~~~q (5)

< 8-1 >

Proposed Table 1 Requirements for Inspection Item B:

Unmitigated butt weld at Once per Cold Leg interval operating Every second B-1 temperature (- Weld Surface I V;"ol (2), (3) -3140 inspection Not 2410) 2: 525°F Fig. 1 Volumetric (4) -3130 period not to I Permissible (274°C) and <  !

exceed 7 yr 580°F (304°C), i (5)

Less than NPS

, 14 (ON 350)

Unmitigated butt weld at

Cold Leg operating Once per temperature (

Weld Surface Visual (2), (3) -31,40 intervat B-2 2410) 2: 525°F Permissible Fig. 1 . Volumetric (4) -3130 I Once per (274°C) and <

interval 580°F (304°C),

NPS 14 or Larger (ON 350) 8.2 Proposed Revision to MRP-139 for Cold Leg Locations Section 6.5.2 should be revised to read "PWSCC Category E welds less than 14 NPS shall be volumetrically inspected 100% every six years. PWSCC Category E welds 14 NPS or greater shall be volumetrically inspected 100% every ten years."

Table 6-1, "Examination Extent and Schedule" for PWSCC Category E shall be revised from "100% every six years" to "100% every six years for less than 14 NPS.

100% every ten years for 14 NPS or greater."

<< 8*2 ~

Section 9: References

1. Materials Reliability Program: Primary System Butt Weld Inspection and Evaluation Guideline (MRP-139, Revision 1), Palo Alto, CA: 2008.

1015009

2. N-770, Revision 1, Alternative Examination Requirements and Acceptance Standards for 1 PWR and Vessel Nozzle Butt Welds nnr:,r/lT."fJ With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application ofListed Mitigation /TrT19n.r1PL Section 1, New York, NY, December 2009.

3. ASME Boiler and Code """',l!\)'"

Inspection ofNuclear Power Plant Components, 2007 LU.LLIV'U Addenda, July 1, 2009.

4. Donavin, Elder, G. and Technical Basis Document for Alloy 82/182 Weld Inspection August 8, 2008.

5. Material Reliability Program: Alloy 82/182 Pipe Butt Weld Safety Assessmentfor US PWR Plant Designs (MRP-113), Palo Alto, CA: August, 2004,1009549.

6. Bamford, W. H. and N. Service with Alloy 600 and Associated Welds in Operating PWRs, Including Repair Activities and Regulatory and Code Actions, of the Fourteenth International Conference on of Materials in Nuclear Power Water Reactors, American Nuclear Society, August 2009.

7. Bamford, W.H. and J" ofAlloy 600 Nozzles and Welds in PWRs: Review of Cracking Events and Repair Experience, Proceedings the Twelfth International Conference on Environmental Degradation of Materials in Power Water American Nuclear Society,

8. PWR Material Reliability Project Interim Alloy 600 Safety Assessments for US PWR Plants (MRP-44), Part 1: Alloy 82/182 Butt Welds, EPRl, Palo Alto, 2001. TP-1001491.

9. Materials Reliability Program: Alloy 82/182 Pipe Butt Weld Safety Assessment for us PWR Plant Westinghouse and CE Plant Designs (MRP-109),

EPRl, Alto, CA: 2005.1009804.

( 9-1 )

10. Materials Reliability Program: Alloy 82/182 Pipe Butt Weld Safety Assessment for US PWR Plant Designs: Babcock & Wilcox Design Plants (MPR-112),

EPRI, Palo Alto, CA: 2004.1009805.

11. Materials Reliability Program: Probabilistic Risk Assessment ofAlloy 82/182 Piping Butt Welds (MRP-116), EPRI, Palo Alto, CA: 2004.1009806.

12. Crack Growth ofAlloy 182 Weld Metal in PWR Environments (MRP-21),

EPRI, Palo Alto, CA: 2000. 1000037.

13. Materials Reliability Program: Crack Growth Ratesfor Evaluating Primary Water Stress Corrosion Cracking (PWSCC) ofAlloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004. 1006696.

14. Flaw Evaluation olCE Design RCP Suction and Discharge Nozzle Dissimilar Metal Welds, Phase III Study, WCAP-17128-NP, Revision 1, Westinghouse, 2010.

15. Material Reliability Program: Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance (MRP-287), EPRI, Palo Alto, CA:

2010. 1021023.

16. Good B., et al, Residual Stress Effects on PWSCC Crack Growth Behavior in Reactor Vessel Nozzle Dissimilar Metal Welds, to appear in Proceedings of ASME Pressure Vessel and Piping Conference,July 2011.

17. U.S. Nuclear Regulatory Commission, An Approachfor Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 1, November 2002.

18. MRP 2010-046, MRP-139, Revision 1 Interim Guidance on Rescission of MRP-139, R1 Mandatory Requirements with Implementation of Code Case N-770;January4,2011

-< 9-2 >

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