ML20133C901

Responds to Press Release of 941006,requesting Action Re Plant,Under 10CFR2.206

ML20133C901
Person / Time
Site:	Peach Bottom
Issue date:	04/03/1996
From:	Russell W NRC (Affiliation Not Assigned)
To:	Ochs R MARYLAND NUCLEAR SAFETY COALITION
Shared Package
ML20112H946	List: DD-96-05, Responds to Press Release Requesting Action Re Plant,Operated by Peco.Director'S Decision DD-96-05 Encl ML20132D090 ML20133C901
References
2.206, DD-96-03, DD-96-3, NUDOCS 9701080138
Download: ML20133C901 (8)

Text

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g k UNITED STATES S NUCLEAR REGULATORY COMMISSION { f WASHINGTON, D.C. 30866-0001

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l g,,* April 3, 1996 Mr. Richard Ochs , Maryland Safe Energy Coalition l l P.O. Box 33111 l Baltimore, MD 21218 i l

Dear Mr. Ochs:

l l I am responding to your press release of October 6,1994, in which the ) Maryland Safe Energy Coalition requested action with regard to Peach Bottom Atomic Power Station of PECO Energy Company. This request was treated as a I petition in accordance with 10 CFR 2.206 of the U.S. Nuclear Regulatory Comission's (NRC's) regulations. i Specifically, the Petition requested that the NRC (1) imediately shut down both reactors at Peach Bottom until the risk of fire near electrical control cables due to combustible insulation is corrected; (2) immediately shut down both reactors at Peach Bottom until all safety class component parts in both l reactor vessels, including the cooling system, the heat transfer system, and the reactor core, are inspected; (3) suspend the Peach Bottom license until an l analysis of the synergistic effects of cracks in multiple parts is conducted; and (4) imediately shut down both reactors at Peach Bottom pending correction of numerous equipment problems identified in certain NRC inspection reports. On December 2,1994, I informed you t' hat I was denying your requests for immediate shutdown and suspension of the operating license, that your Petition was being evaluated under 10 CFR 2.206 of the Comission's regulations, and that action would be taken in a reasonable time. The staff has completed its review of your request to imediately shut down both reactors at Peach Bottom until the risk of fire near electrical control cables due to combustible insulation is corrected. For the reasons stated in the enclosed Director's Decision (DD-96-03), your request is denied. The remaining issues that you raised that were the basis for Requests 2, 3, and 4 of your Petition of October 6,1994, are still under consideration by the NRC staff. A Final Director's Decision will be issued upon completion of the NRC staff's review. A copy of the enclosed Decision will be filed with the Secretary of the Commission for the Commission to review in accordance with 10 CFR 2.206(c). As provided by this regulation, the Decision will constitute the final action of the Commission 25 days after the date of issuance of the Decision unless l the Comission, on its own motion, institutes a review of the Decision within that time. The documents cited in the enclosed Decision are available for review at the Comission's Public Document Room, Gelman Building, 2120 L Street, N.W., Washington, D.C., and at the local Public Document Room for the named facilities. l 9701080138 960610 {DR ADOCKOS00g7

R. Ochs April 3, 1996 I have also enclosed a copy of the notice, " Issuance of Director's Decision Under 10 CFR 2.206," which includes the complete text of DD-96-03, that is being filed with the Office of the Federal Register for publication.  ! l Sincerely, , l O i William T. Russell, Director Office of Nuclear Reactor Regulation Docket Nos.: 50-277 and 50-278 (10 CFR 2.206)

Enclosures:

1. Director's Decision

2. Federal Reaister Notice cc w/ encl: See next page 1

4 R. Ochs April 3, 1996 I have also enclosed a cosy of the notice, " Issuance of Director's Decision ' Under 10 CFR 2.206," whic1 includes the complete text of DD-9643, that is being filed with the Office of the Federal Register for publication. Sincerely, Original Signed By l eILLIAM T. RUSSELU William T. Russell, Dir ctor Office of Nuclear Reactor Regulation Docket Nos.: 50-277 and 50-278 (10 CFR 2.206)

Enclosures:

1. Director's Decision

2. Federal Reoister Notice cc w/ encl: See next page DISTRIBUTION: See attached list

• Previous Concurrence Office PDI-2/ll,, PD 2/PM EDI-2/PD TECH ED
• SPl,R/C M EELB/C Name MG/Edi[ JSNek[ Ntdz BCalure CbNn JCalvo [

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_. . - . - . . - - _ _ _ - . - - . - - - - -.~-- -._ .- - - . _ - . . _ . - l'. PECO Energy Company Peach Botto2 Ator.ic Powsr Station,

       ,                                                                          Units 2 and 3 i

cc: i 1

J. W. Durham, Sr., Esquire Mr. Rich R. Janati, Chief Sr. V.P. & General Counsel Division of Nuclear Safety i PECO Energy Company Pennsylvania Department of 2301 Market Street, S26-1 Environmental Resources 4

ohiladelphia, Pennsylvania 19101 P. 0. Box 8469 Harrisburg, Pennsylvania 17105-8469 l PECO Energy Company , ATTN: Mr. G. R. Rainey, Vice President Board of Superviscrs

Peach Bottom Atomic Power Station Peach Bottom Township

Route 1. Box 208 R. D. #1 Delta, Pennsylvania 17314 Delta, Pennsylvania 17314 l

l PECO Energy Company Public Service Comission of Maryland ATTN: Regulatory Engineer, A4-5S Engineering Division Peach Bottom Atomic Power Station Chief Engineer ! Route 1, Box 208 6 St. Paul Centre j Delta, Pennsylvania 17314 Baltimore, MD 21202-6806 i Resident Inspector Mr. Richard McLean i U.S. Nuclear Regulatory Comission . Power Plant and Environmental Peach Bottom Atomic Power Station Review Division i P.O. Box 399 Department of Natural Resources i Delta, Pennsylvania 17314 B-3, Tawes State Office Building ' Annapolis, Maryland 21401 Regional Administrator, Region I l U.S. Nuclear Regulatory Comission Dr. Judith Johnsrud 475 Allendale Road National Energy. Comittee

King of Prussia, Pennsylvania 19406 Sierra Club 1 433 Orlando Avenue

) Mr. Roland Fletcher State College, PA 16803

Department of Environment i

201 West Preston Street Bryan W. Gorman, Manager Baltimore, Maryland 21201 Joint Owners / External Affairs Interface Public Service Electric and Gas } A. F. Kirby, III Company 3 External Operations - Nuclear P.O. Box 236 Delmarva Power & Light Company Hancocks Bridge, NJ 08038-0236 P.O. Box 231 , Wilmington, DE 19899 PECO Energy Company G. D. Edwards, Plant Manager l j Mr. George A. Hunger, Jr. Peach Bottom Atomic Power Station '

!            Director-Licensing, MC 62A-1                                       Route 1, Box 208 PECO Energy Company                                                Delta, PA 17314                                     l
;            Nuclear Group Headquarters                                                                                             1 4             Correspondence Control Desk l            P.O. Box No. 195 Hancocks Bridge, NJ 08038                                                                                              l l

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00-96-03 UNITED STATES OF AMERICA NUCLEAR REGULATORY ColeilSSION OFFICE OF NUCLEAR REACTOR REGULATION William T. Russell, Director In the Matter of )

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All Reactor Licensees With Installed ) Thermo-Lag Fire Barrier Material. ) DIRECTOR'S DECISION UNDER 10 CFR 2.206 l l I. INTRODUCTION l By letter dated September 26, 1994, the citizens for Fair Utility Regulation and the Nuclear Information and Resource Service (NIRS); by press release dated October 6,1994, the Maryland Safe Energy Coalition; by separate l letters dated October 21, 1994, the GE Stockholders' Alliance and Dr. D. K. Cinquemani; by letter dated October 25, 1994, the Toledo Coalition for Safe Energy; by letter dated October 26, 1994, R. Benjan; by letter dated November 14, 1994, B. DeBolt; and by letter dated December 8,1994, NIRS and the Oyster Creek Nuclear Watch (the Petitioners), requested that the U.S. Nuclear Regulatory Commission (NRC) take action with regard to the use of Thermo-Lag by reactor licensees and that their letters be treated as Petitions pursuant to Section 2.206 of Title 10 of the Code of Federal Reaulations (10 CFR 2.206). The Citizens for Fair Utility Regulation and NIRS requested that (1) Texas Utilities Electric Company (TU Electric), licensee of Comanche Peak Steam Electric Station, Unit 1, perform additional destructive. analysis for Thermo-Lag configurations in proportion to the total installed amount of Thermo-Lag to determine the degree of " dry joint" occurrence, (2) the licensee perform fire tests on upgraded " dry joint" Thermo-Lag configurations for conduit and cable trays to rate the barrier as a tested configuration in l l

l . l - 2 compliance with fire protection regulations, and (3) the NRC inmediately suspend the Comanche Peak Unit 1 license until the above corrective actions are taken. The Maryland Safe Energy Coalition requested immediate shutdown of both reactors at the Peach Bottom plant until the risk of fire near electrical control cables due to combustible insulation is corrected.' Dr. Cinquemani and the Toledo Coalition for Safe Energy requested that the NRC insediately shut down all reactors where Thermo-Lag is used until it has been removed and replaced. The GE Stockholders' Alliance requested shutdown of all reactors i l where Thermo-Lag is used until it has been removed and replaced with fire-retardant material meeting NRC standards. R. Benjan requested immediate shutdown of all reactors where Thermo-Lag is used. B. DeBolt requested shutdown of all reactors in which Thermo-Lag is used until it has been removed and replaced. NIRS and the Oyster Creek Nuclear Watch requested that NRC immediately suspend GPU Nuclear Corporation's (GPUN's) operating license for Oyster Creek Nuclear Generating Station (OCNGS) until GPUN removes Thermo-Lag fire barrier material and replaces it with a competitive product that meets current NRC fire protection regulations. As a basis for their requests concerning Thermo-Lag 330-1 fire barrier upgrades, the Citizens for Fair Utility Regulation and NIRS Petitioners stated that (1) the licensee's records on the original installation of Thermo-Lag fire barriers on conduits and cable trays indicate that its contractor followed specifications for pre-buttering all joints; (2) NRC Inspection Reports 50-455/93-42 and 50-446/93-42 found, based on destructive analysis l l

            'Thu Petition submitted by the Maryland Safe Energy Coalition expressed several concerns in addition to the fire hazard issue. These other issues, l     that is other than the fire hazard issue, will be the subject of a separate Director's Decision.

l

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i i'. 3 documents, that a concern did exist where Thermo-Lag conduit joints fell apart easily and did not appear to have any residual material of a buttered surface, I i indicative of a joint that had not been pre-buttered; (3) the " dry joint" j deficiency appeared in Room 115A and other areas of the unit; (4) the licensee i directly contradicts an NRC inspector's findings that were determined in part ) by destructive analysis; (5) the " dry joint" or absence of pre-buttering of

Thermo-Lag panels can be determined only by destructive analysis and cannot be i

1 determined by a walkdown visual inspection; (6) the findings reported in the t Comanche Peak Unit 1 Region IV Inspection Reports 50-455/93-42 and 50-446/93-42, based on the limited amount of destructive analysis conducted at i the unit, constitute a substantial documentation of installation deficiencies i =

found in Thermo-Lag fire barriers as documented in NRC Information Notice 2

(IN) 91-79, " Deficiencies in the Procedures for Installing Thermo-Lag Fire j Barrier Materials," December 6, 1991, and IN 91-79, Supplement 1,

          " Deficiencies Found in Thermo-Lag Fire Barrier Installation," August 4, 1994;               I (7) neither the NRC nor the industry, by its agent Nuclear Energy Institute

(NEI), nor a utility, have conducted fire tests on dry-fitted or " dry joint" 1

j upgraded configurations of Thermo-Lag 330-1; and (8) the presence of " dry ! joint" upgraded configurations in Comanche Peak Unit I constitutes an untested

application of Thermo-Lag fire barriers.

l As a basis for the requests concerning Thermo-Lag 330-1 fire barrier i upgrades, the Maryland Safe Energy Coalition stated that the manufacturer of i i the flame retardant (Thermo-Lag insulation) was indicted on criminal charges (of falsifying tests of the effectiveness of the insulation as a fire 1 barrier), and fire near the electrical control cables, due to combustible

!        Thermo-Lag insulation, could cause a catastrophic meltdown.
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4 As the bases for their requests, Dr. Cinquemani, the Toledo Coalition for Safe Energy, the GE Stockholders' Alliance, and R. Benjan stated either individually or collectively that (1) the widespread use of Thermo-Lag in more l than 70 reactors presents a safety crisis; (2) the NRC has known since 1982 ' l that Therme-Lag fails NRC performance standards for material that protects i i vital electrical cables for ampacity rating and fire resistance; (3) Thermo-Lag has failed not only NRC tests, but almost all other independent tests; (4) Thermo-Lag is combustible, contrary to NRC regulations, and is an ineffective fire barrier; (5) the use of Thermo-Lag could lead to shorts, to failure of the cables in an emergency, and to fire; (6) Thermo-Lag is faulty in that fraudulent ampacity ratings allowed utilities to use smaller cable j than permitted by design requirements, causing the cable to overheat and its I insulation to deteriorate; (7) the NRC has stated that fire at some nuclear power plants can contribute as much as 50 percent of the risk to a core meltdown, and a typical reactor will have three to four significant fires during its licensed lifetime; (8) Thermal Science, Inc. (TSI), the manufacturer of Thermo-Lag, and its President were indicted by a Federal grand jury on seven criminal charges related to conspiracy to defraud the U.S. Government in regard to the effectiveness of Thermo-Lag; and (9) the hourly fire watches at the Davis-Besse Nuclear Power Plant operated by Toledo Edison da not replace fire barrier material and do not prevent fires. As the bases for his request, B. DeBolt stated that Thermo-Lag fails to meet NRC regulations concerning combustibility and that the manufacturer of Thermo-Lag was indicted for defrauding the Government and the utilities. l Among the many bases for their request, NIRS and the Oyster Creek Nuclear Watch stated that (1) Southwest Research Institute (SwRI) conducted fire tests

4 5 - on Thermo-Lag 330-1 specimens for GPUN and reported that all specimens ignited approximately 2 seconds after it was inserted into the furnace and failed i specified criteria because of flaming after the first 30 seconds of testing, an outside temperature rise higher than 30 'C, and a weight loss of 50 percent; (2) GPUN's operation of OCNGS with knowledge of the SwRI report is an example of GPUN's reckless disregard for fire protection and public-safety; (3) in the event of fire, Thermo-Lag is likely to fail its intended function of protecting vital electrical cables running from the control room to plant safety systems used to shut down the reactor; (4) current installations of Thermo-Lag are likely to fail in less time than I hour (when smoke detectors and automatic sprinkler systems are present) or 3 hours (when there are no fire detection and suppression systems) that NRC regulations require for fire barriers to withstand fire; (5) the NRC Inspector General issued a report in . August 1992 condemn'ing NRC's handling of the Thermo-Lag issue and documenting the NRC staff's failure to understand the scope of the problem; (6) in April 1994, Industrial Testing Laboratories and its President pleaded guilty to five felony counts of aiding and abetting the distribution of falsified test data; (7) on September 29, 1994, the U.S. Department of Justice issued a seven-count indictment against the manufacturer of Thermo-Lag and its Chief Executive Officer for willful violations of the Atomic Energy Act, conspiracy to conceal material facts, and making false statements to defraud the United States in connection with $58 million in fire barrier material; (8) GPUN has known since at least August 11, 1992, that Thermo-Lag 330-1 as a structural base material is combustible and that GPUN was in violation of Appendices A and R to 10 CFR Part 50 and the NRC Standard Review Plan, NUREG-0800; (9) GPUN failed to report the SwRI test results in response to a request for additional

a. 6 infonsation regarding Gen:~ic Letter (GL) 92-08 ("Thermo-Lag 330-1 Fire Barriers") of February 10, 1994, when asked to describ the Themo-Lag 330-1 fire barriers installed as required to meet 10 CFR Part 50, Appendix R; and u (10) continued reliance on fire watches at OCNGS is an unreasonable and unnecessary hazard to the public health and safety because of an inoperable fire protection system for safe shutdown of the reactor and installed combustible material on the shutdown systems. . On November 7,1994, I informed the Citizens for Fair Utility Regulation and NIRS that the request for an imediate suspension of the Comanche Peak l Unit 1 operating license was denied. On December 2,1994, I informed t.he Maryland Safe Energy Coalition that the request for an immediate shutdown of l the Peach Bottom plant and for an imediate suspension of the Peach Bottom license was denied. On December 15, 1994, I informed the GE Stockholders Alliance, Dr. D. K. Cinquemani, the Toledo Coalition for Safe Energy, and R. Benjan that the imediate suspension of the operating licenses of all reactors where Thermo-Lag is used was denied. On January 3, 1995, I informed NIRS and the Oyster Creek Nuclear Watch that the imediate suspension of the OCNGS operating license was denied. On January 19, 1995, I informed B. DeBolt that the request for immediate suspension of the operating licenses of all reactors in which Themo-Lag is used was denied. The decisions were based on the following: (1) the staff is addressing deficiencies in fire barriers constructed with Thermo-Lag material as part of a Comission-approved action plan and has issued several bulletins and a generic letter to the nuclear industry to provide information and guidance, (2) fire barrier systems constructed with Thermo-Lag have been identified and declared inoperable, and (3) compensatory measures (fire watches) approved by the NRC have been l

4; i 4 7 j instituted. Additionally in the above correspondence, all Petitioners were j informed that the Petitions were being treated pursuant to 10 CFR 2.206 and l had been referred to this office for action pursuant to 10 CFR 2.206 of the ! Commission's regulations and that appropriate action would be taken within a i i reasonable time. ] For the reasons stated below, the Petitions have been denied.

II. PACKGROUND

] The picture painted by the Petitioners of inaction by the NRC staff in j responding to the issues presented by the use of Thermo-Lag is at odds with 4 the facts. A review of the chronological development of the issues shows that j the NRC staff has been working diligently to resolve the issues and has j consistently sought to ensure that there is adequate protection of the public 1 j health and safety. It is also inaccurate to contend that Thermo-Lag generic j deficiencies have been known since 1982. As can be seen from the following j information, the development of the Thermo-Lag issue has been evolutionary. 4 j Reports of problems regarding Thermo-Lag began to surface in the late 1980s L when Gulf States Utilities, the licensee for River Bend Station, discovered j some cracks and wear damage due to installation deficiencies (Licensee Event Report 87-005, March 25, 1987) and declared the material inoperable as a fire l barrier. The licensee further discovered that stress skin was missing on all ! 3-hour Thermo-Lag fire barriers in the turbine building as a result of an i ! installation error. In a series of plant-specific tests performed by Gulf 1 j States Utilities in 1989, Thermo-Lag barriers failed to meet the fire ,1 endurance test acceptance criteria. Gulf States Utilities categorized all 1

1-hour and 3-hour barriers as indeterminate and implemented compensatory j measures in the form of fire watches. Other isolated plant-specific fire l

i i

8 protection problems had been found during NRC inspections at various utilities as early ar, 1982 and had been acted on by the NRC staff. These problems were treated as plant-specific issues and were not considered as indications of generic problems. In February 1991, the NRC received' allegations that Thenno-Lag did not provide fire protection for electrical cables as claimed by the vend 6C In response, in May 1991, the NRC visited River Bend Station to review the installation procedures and the failed fire endurance tests and concluded that a generic concern existed with 30-inch-wide cable trays. The NRC alerted the industry of the results of the test failures in IN 91-47, " Failure of Thermo-Lag Fire Barrier Material To Pass Fire Endurlince Test," August 6,1991. In June 1991, the Office of Nuclear Reactor Regulation (NRR) established a special review team to investigate the safety significance and generic -

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applicability of technical issues regarding all e ations and operating experience concerning Thermo-Lag fire barriers. In its final report, which was issued with IN 92-46, "Thermo-Lag Fire Barrier Naterial Special Review Team Final Report Findings, Current Fire Endurance Testing, and Ampacity Calculation Errors," June 23, 1992, the special review team reached the following conclusions: The fire-resistive ratings and the ampacity derating factors for the Therno-Lag fire barrier system were indeterminate. Some licensees had not reviewed and evaluated the fire endurance test results and the ampacity derating test results used as the licensing basis for their Thermo-Lag barriers to determine the validity of the tests and the applicability of the test results to their plant designs. 1 l l I

? 9 Some licensees had not reviewed the Thermo-Lag fire barriers installed in their plants to ensure that they met NRC requirements and guidance, i i such as that provided in GL 66-10 " Implementation of Fire Protection ] Requirements," April 24, 1986. Some licensees used inadequate or' incomplete installation procedures during the construction of their Thermo-Lag barriers. ., After the special review team completed its charter, the NRC staff j prepared an action plan that provided a process to resolve technical issues f identified with Thermo-Lag fire barrier systems. The NEI, formerly the Nuclear Management and Resources Council (NUMARC), agreed to coordinate industry efforts to resolve the issues. In regard to the Petitioners' allegations of NRC's inaction in responding to the issues presented by the use of Thermo-Lag, the significant progress made by the NRC staff and the nuclear reactor licensees in resolving Thermo-Lag issues speaks to the contrary. The NRC staff has issued a number of generic comunications related to Thermo-Lag, which include the following: (1) two bulletins: BUL 92-01, " Failure of Thermo-Lag 330 Fire Barrier System To Maintain Cabling in Wide Cable Trays and Small Conduits Free From Fire Damage," June 24, 1992, and BUL 92-01, Supplement 1, " Failure of Thermo-Lag 330 Fire Barrier System To Perform Its Specified Fire Endurance Function," August 28, 1992; (2) two generic letters: GL 92-08, "Thermo-Lag 330-1 Fire Bar.lers," December 17, 1992, and GL 86-10, Supplement 1, " Fire Endurance Test Acceptrnce Criteria for Fire Barrier Systems Used To Separate Redundant Safe Shutdown Trains Within the same Fire Area," March 25, 1994; and (3) 12 information notices: IN 91-47; IN 91-79; IN 91-79, Supplement 1; IN 92-46; IN 92-55, " Current Fire Endurance Test Results for Thermo-Lag Fire Barrier

g 10 Material," July 27, 1992; IN 92-82, "Results of Thermo-Lag 330-1 Combustibility Testing," December 15, 1992; IN 94-22 " Fire Endurance and Ampacity Derating Test Results for 3-Hour Fire-Rated Therno-Lag 330-1 Fire Barriers," March 16, 1994; IN 94-86, " Legal Actions Against Thermal Science, Inc., Manufacturer of Thermo-Lag," December 22, 1994; IN 95-27, "NRC Review of Nuclear Energy Institute, Thermo-Lag 330-1 Combustibility Evaluation Methodology Plant Screening Guide," May 31, 1995; IN 95-32, "Thermo-Lag 330-1 Flame Spread Test Results," August 10, 1995; IN 95-49, " Seismic Adequacy of Thermo-Lag Panels," October 27, 1995, and IN 94-86, Supplement 1, " Legal Actions Against Thermal Science, Inc., Manufacturer of Thermo-Lag," , l November 15, 1995. ' The NRC staff, the nuclear industry, and others have expended much time and many resources to address and resolve the Thermo-Lag issues. The NRC staff developed comprehensive fire test guidance and acceptance criteria and worked with industry to improve existing ampacity test procedures. The NRC starf and industry performed about 100 fire endurance and ampacity derating tests of Thermo-Lag fire barrier materials and full-scale test assemblies. The fire endurance tests established the limitations and the true fire-resistive capabilities of certain Thermo-Lag fire barrier configurations, without relying on the fire endurance test data supplied by TSI, the j manufacturer of Therwo-Lag. On the basis of some of these tests, the NRC  ; staff concluded that existing Thermo-Lag barriers could be upgraded with some additional Therno-Lag material to satisfy NRC regulations. Precluding all use of Thermo-Lag materials for current and future fire barrier installations would remove a realistic option for resolving safety issues. Therefore, the NRC staff does not object to the use of Thermo-Lag in specific applications,

e 11 where, through upgrades, NRC requirements are satisfied. The NRC staff issued three requests for additional information (RAls) regarding GL 92-08 to each - licensee using Thermo-Lag to obtain information on the specific Thermo-Lag material installed at each plant. The NRC staff reviewed and approved g comprehensive Thermo-Lag fire barrier programs proposed by TU Electric for

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I Comanche Peak Steam Electric Station, Unit 2, and by Tennessee Valley j Authority (TVA) for Watts Bar Nuclear Power Plant, Unit 1, which attests to the fact that Thermo-Lag barriers can meet NRC fire protection guidelines and i l requirements. The NRC staff completed toxicity tests of Thermo-Lag material, i The NRC staff and the industry completed chemical composition, combustibility, and flame spread tests of Thermo-Lag materials. Finally, the NRC staff reassessed previous technical conclusions to determine the extent to which the NRC staff and industry relied on information supplied by TSI to reach these conclusions. The staff had concerns about the reliability of information and data supplied by TSI that have been or could be used to make judgments regarding Thermo-Lag materials. The NRC staff identified and categorized the

issues and previous conclusions and used the results of the industry-wide a

testing program regarding the chemical composition of Thermo-Lag, as discussed below, to determine if the in-plant Thermo-Lag materials were consistent. The , results of this reassessment indicated that previous technical conclusions l were valid independent of the information provided by TSI. The staff therefore concluded that additional action to reassess the issues or reverify the previous conclusions was not needed. The NEI testing program on the chemical composition of Thermo-Lag analyzed samples from 18 utilities representing 25 nuclear power plants. The samples represented Thermo-Lag material manufactured between 1984 and 1995.

12 NEI performed pyrolysis gas chromatography evaluation of 169 samples to assess organic chemical composition and performed energy-dispersive X-ray spectroscopy of 33 samples to assess inorganic chemical ' composition. On the basis of the tests, NEI concluded that (1) all of the samples contained the constituents identified by TSI as essent'ial to fire barrier performance; (2) the composition of the samples was consistent; and (3) the test: results provided a basis on which to close NRC questions about chemical composition and product consistency and for utility use of generic test data relative to fire endurance ratings, flame spread, heat release, ampacity derating, and other material properties. The NRC staff test program on the chemical composition of Thermo-Lag was conducted by the National Institute of Standards and Technology (NIST) during 1992 and 1995. NIST analyzed 21 samples that were either collected by the staff during site visits to plants and test laboratories or provided by TVA, Gulf States Utilities, Commonwealth Edison Company, and NEI. The analysis included elemental and amonia analysis, pyrolysis, gas chromatography, mass spectrometry, and X-ray fluorescence. These analytical techniques indicated that all of the samples were similar in their bulk chemical composition. Thes.e results were consistent with the results of the NEI chemical testing program pertaining to the chemical composition and uniformity of Thermo-Lag. Industry-wide progress has generally been comensurate with the complexity of the plant-specific. issues and the amounts of Thermo-Lag installed at the individual plants. Several licensees have initiated programs to replace Thermo-Lag and are performing plant-specific tests of other fire barrier materials such as Mecatiss (Florida Power & Light for Crystal River , Unit 3) and Darmatt KM-1 (Carolina Power & Light for Brunswick, IES Utilities,

13 Inc., for Duane Arnold Energy Center, Commonwealth Edison Company for LaSalle County Station, and Northern States Power Company for Prairie Island Nuclear Generating Plant). The NRC staff is reviewing the plant-specific fire endurance test programs and has recently approved the plant-specific application of Darmatt KM-1 fire barrier at the LaSalle plant. The remaining licensees have submitted to the NRC staff detailed plans and schedules for resolving the issues at their plants. Most licensees are pursuing a combination of such options as upgrading existing Thermo-Lag fire barriers to meet NRC fire barrier requirements, replacing Thermo-Lag fire barriers with another type of fire barrier, reducing or eliminating reliance on Thermo-lag fire barriers by relocating equipment and cables and by post-fire safe-shutdown reanalysis, installing additional fire protection features such as automatic sprinkler systems, and requesting configuration-specific exemptions when such exemptions are allowed by NRC regulations and are technically justified to provide a level of safety equivalent to that prescribed by the regulations. The NRC staff has completed its review of the plans for resolving fire protection issues that were proposed by most of the licensees. As with any issues as technically complex, challenging, and resource intensive as those presented by Thermo-Lag barriers, some plant-specific questions remain. However, the number of issues has steadily declined. The NRC staff and the licensees will continue to address the residual questions on a case-by-case basis as they arise, and the NRC staff will continue to follow up with individual licensees on their corrective actions, as appropriate. Every licensee with Thermo-Lag fire barriers will continue to maintain NRC-approved compensatory measures, such as fire watches, until its permanent corrective

 .                                                                                     ,4 14 actions are implemented. Therefore, the public health and safety are protected.

The NRC's " defense-in-depth" fire protection concept relies on protecting safe shutdown functions by achieving a balance among three echelons , or levels of protection, which are (1) fire prevention activities; (2) the ability to rapidly detect, control, and suppress a fire; and (3) physical separation of redundant safe shutdown functions. Weaknesses found in one area may be dealt with by enhancing the protection capabilities of the remaining areas.2 The NRC foresaw cases in which fire protection features would be inoperable and required licensees, through technical specifications or approved fire protection plans controlled by license conditions, to provide compensation for the deficient condition. The concept of allowing alternative actions to compensate for an inoperable condition or component is used in various programs associated with the operation of nuclear power plants and has long been an integral part of NRC regulatory requirements.3 ' The fire endurance test results contained in NRC BUL 92-01 and NRC BUL 92-01, Supplement 1, confirmed that certain Thermo-Lag fire barrier configurations compromise one facet of the fire protection defense-in-depth concept. In response to NRC BUL 92-01 and its supplement, the licensees for plants using Thenno-Lag fire barriers established fire watches in accordance with their technical specifications or license conditions as a compensatory 2 The " defense-in-depth" concept is detailed in the "NRC Standard Review Plan," NUREG-0800, Section 9.5.1, " Fire Protection Program," page 9.5.1-10. 1 3 NRC GL 91-18, "Information to Licensees Regarding Two NRC Manual Sections on Resolution of Degraded and Nonconforming Conditions and 1 Operability," issued November 7,1991, and NRC Inspection Manual, Part 9900, j

   " Resolution of Degraded and Nonconforming Conditions," issued October 31,             j 1991.                                                                                   i l

l j

                                                                                     \
 ,                                            15 measure. Fire watches are personnel trained by the licensees to inspect for the control of ignition sources, fire hazard!,, and combustible materials; to
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look for signs of incipient fires; to provide prompt notification of fire hazards and fires; and to take appropriate actions to begin fire suppression activities. Generally, therefore, by providing additional fire prevention activities through enhanced detection capabilities to find fire hazards and in the case of a fire, augmented suppression activities before a barrier's ability to endure a fire is challenged, fire watches compensate for degraded fire barriers. The NRC staff has carefully evaluated the issues associated with continued use of Thermo-Lag material, including the use of fire watches to compensate for any degradation in the effectiveness of required fire barriers. Such compensatory actions provide an adequate level of fire protection without i 1 an undue risk to the health and safety of the public. Licensees have i established fire watches to compensate for degraded and possibly inoperable fire barriers. Also, licensees rely on a defense-in-depth concept that incorporates multiple safety measures. Automatic fire detection and suppression systems are provided in most areas that have safe shutdown equipment. Trained fire brigades are required 24 hours a day at all plants. All areas that have safe shutdown equipment have manual fire suppression features. Fuels that can feed a fire and ignition sources to start a fire are controlled. The combination of fire watches and the defense-in-depth fire protection features provides an adequate level of fire protection until licensees implement permanent corrective actions.

sq

!.                                                                                                           16 l                                                  Taken together, these factors represent an adequate means of fire protection at the plants using Thermo-Lag to ensure, with margin,' that j

) operation can be conducted without an undue risk to the health and safety of the public. Nevertheless, with these considerations in mind, the NRC staff addressed below the Petitioners' specific concerns to demonstrate that no substantial health and safety issue has been raised. III. RESPONSE TO SPECIFIC CONCERNS [ . { The Petitioners alleged that (1) the NRC has been slow to enforce its 1 i 1 own regulations, (2) fire watches do not replace fire barriers and continued l 1 reliance on fire watches is an unreasonable and unnecessary hazard to the l public health and safety because of an inoperable fire protection system for 1 l safe shutdown of the reactor and installed combustible material on the

l. shutdown systems, (3) utilities are in violation of NRC requirements because  !

J i Thermo-Lag is combustille and could contribute to a fire instead of protecting 1 l from it, and, in spite of the danger, the NRC allows continued use of Thermo-Lag (4) faulty ampacity ratings could result in the use of j inappropriate cables, which, if undersized, could overheat and cause its insulation to deteriorate, (5) the licensee for Oyster Creek did not report to the NRC its findings regarding the c.ombustibility of Thermo-Lag and, (6) the l Thermo-Lag barriers have been improperly installed at Comanche Peak Unit 1, I which contributes further to the poor performance of Thermo-Lag. 2 l The NRC staff acknowledged and has stated that certain Thermo-Lag fire j barrier configurations have failed to demonstrate the ability to perform their fire resistance functions. In this regard, the NRC staff, in BUL 92-01, j 'The fact that Thermo-Lag barriers, as installed, will provide protection

for some period of time is supported by, among others, the fire endurance test

results documented in IN 92-55.

t

   . _   _ _ _ _ _ . _ .       _. --         _ _ _ _ _ _ - _ . _ _      _ _ _    . . . - . - - - . . - . . - .. .---.~ . _ . - _ _ _

1 17 Supplement 1, has stated that Thermo-Lag fire barriers should be treated as I inoperable until licensees can declare the fire barriers operable on the basis of successful, applicable tests. Given the foregoing deficiencies identified for Thermo-Lag, the NRC staff concluded that compensatory measures are necessary until a licensee can declare fire barriers operable on the basis of applicable tests that demonstrate successful barrier performance. ) 1 The Petitioners also asserted that (1) the NRC should have protected the public and not Rubin Feldman, the President of the company manufacturing Thermo-Lag, and (2) public safety has been compromised by NRC's seeming complicity with utilities.5 l A. Reaulatory Comoliance The NRC staff acknowledges that certain fire endurance tests have demonstrated that Thermo-Lag barriers may not meet the fire endurance rating criteria set forth in Section I!!.G.. of Appendix R to 10 CFR Part 50. This acknowledgment does not mean, however, that there no longer is reasonable assurance of protection of the public health and safety or that such actions as the shutdown of all reactors using Thermo-Lag and the suspension of Comanche Peak, Peach Bottom, and Oyster Creek operating licenses are warranted. It should first be noted that Appendix R, which sets forth criteria for specific fire protection features to protect safe shutdown systems, is applicable only to facilities that commenced operation prior to 1979. Facilities commencing operation on or after January 1, 1979, although not

                                  'These statements could be interpreted as the appearance of unwarranted favoritism toward the manufacturer of Thermo-Lag and complicity with utilities. Therefore, the Petitions were referred to the NRC Office of the Inspector General.

l ' 18 bound by Appendix R, generally are bound by licensing commitments to follow the criteria set forth in Appendix R through license conditions.' ' Even assuming that all of the plants in which Thermo-Lag is installed and that conmienced operation prior to 1979 are not in compliance with Appendix R, it does not follow that the failure to comply with a regulation indicates the absence of adequate protection. The Commission has explained l that-- l [W)hile it is true that compliance with all NRC regulations ' provides reasonable assurance of adequate protection of the public health and safety, the converse is not correct, that failure to comply with one regulation or another is an indication of the absence of adequate protection, at least in a situation where the Commission has reviewed the noncompliance and found that it does l not pose an " undue risk" to the public health and safety. l l (Ohio Citizens for Responsible Energy, DPRM 88-4, 28 NRC 411 (1988).) All the plants using Thermo-Lag have instituted fire watches as required by their action statements regarding inoperable barriers contained in their technical specifications or fire protection programs subject to license conditions. Generally, action statements provide alternative remedial actions to shutting down a plant when limiting conditions for operation are not met. Compliance with the required remedial actions provides reasonable assurance that the public health and safety is protected notwithstanding the plant's i i continued operation and its failure to meet the respective limiting condition ' for operation. Here, since all of the plants using Thermo-Lag have implemented the required fire watches in accordance with plant-specific 1

                          'In addition, there are a very limited number of plants which commenced operation on or after January 1,1979, that are not subject to specific license conditions but whose licensees have made commitments to comply with l               NRC fire protection requirements, including Section !!I.G. of Appendix R.                                The NRC is elevating these commitments to license conditions.

                                                                                                                                                     .R
    ,                                                                                                   lg         -

requirements, their continued operation does not pose an undue risk to the public health and safety. The Petitioners assert that fire watches do not replace fire barriers and continued reliance on fire watches is a hazard to public safety. The NRC staff acknowledges that fire watches do not replace fire barriers. However, as will be discussed in greater detail later in this Decision, firrwatches are judged by the NRC to be acceptable compensatory measures and are legally sanctioned remedial actions based on 10 CFR 50.36(c)(2).7 In sum, notwithstanding the failure to have operable fire barriers ! meeting the fire endurance rating criteria specified by Section III.G. of l Appendix R, a plant is not necessarily unsafe to continue operation. To the contrary, fire watches are judged by the NRC to be adequate remedial measures that provide reasonable assurance that the public health and safety is - protected. By'rea' son of compliance tiy all facilities using Thermo-Lag with their technical specifications or fire protection program action statements requiring the implementation of fire watches, protection of the public health a:.

                              . safety is still reasonably ensured for such plants.                                    Because the Commission has discretion regarding enforcement of its regulations, and given the circumstances here in which no significant health and safety issues have been raised, enforcement action of the nature requested by the Petitioners is not warranted.

7 I In instances in which fire protection programs have been moved from technical specifications and are now subject to license conditions, the NRC's t ' approval of the fire protection programs subject to license conditions provides the legal basis for the implementation of fire watches as a remedial measure.

j 20 B. Ability of Fire Watches to Comnensate for a Dearaded Barrier One of the Petitioners' allegations is that the measures taken by l licensees to compensate for degraded barrier conditions, specifically fire watches, are not adequate to protect the public health and safety. The Petitioners have questioned the continued reliance on fire watches in the l light of an inoperable fire protection system for safe plant shutdown and the combustibility of Thermo,-Lag. In addition, the Petitioners claim that a fire watch does not replace a fire barrier in that fire watches are not preventive. Despite the acknowledged shortcomings identified with certain Thermo-Lag fire barriers and after fully considering the arguments presented by the Petitioners regarding the ability of fire watches to provide adequate compensation, the NRC staff has determined that compensatory measures using fire watches are adequate and acceptable to ensure public health and safety until permanent corrective measures are implemented. The use of fire watches in instances of degraded or inoperable barriers is an integral part of NRC-approved fire protection programs. In general, these NRC staff-approved compensatory measures specify the establishment of a continuous fire watch or an hourly fire watch in cases in which automatic detection systems protect the affected components. Although it is true that Thermo-Lag is intended as a barrier and fire watch personnel cannot act as physical shields, a fire watch provides more than simply a detection function. Personnel assigned to fire watches are trained by the licensee to inspect for the control of ignition sources, fire hazards, anti combustible materials; to look for signs of incipient fires; to provide prompt notification of fire hazards and fires; and to take appropriate action to begin fire suppression activities. Fire watch personnel are capable of determining the size, the I i i

4 21 r actual location, the source, and the type of fire--valuable information that cannot be provided by an automatic fire detection system. During a plant fire, compartment temperatures are likely to be less severe at the early stages. On the basis of enhanced capabilities provided by fire watches and notwithstanding that the level of barrier-type protection may be reduced, the NRC staff has determined that there is an adequate margin of safety to ensure protection in cases in which fire watches are approved. The goal of the NRC staff's Thermo-Lag Action Plan is directed towards restoring the functional capability of fire barriers as soon as practicable. There is not a time limit associated with the use of fire watches as a compensatory measure. Given the margin of safety a fire watch brings to a fire protection program, as discussed above, the NRC staff has determined that continuing the use of fire watches while barriers are inoperable is acceptable. However, the NRC believes that notwithstanding interim reliance on compensatory measures, appropriate actions must be taken by licensees to restore operability of Thermo-Lag barriers. Individual licensees have provided schedules for restoring operability and these are being tracked by the NRC staff. l The NRC staff has carefully evaluated the use of fire watches to compensate for any degradation in the effectiveness of required fire barriers and has concluded that fire watches continue to ensure protection of the public health and safety. Therefore, the Petitioners' assertion that the measures taken by licensees to compensate for degraded fire barrier conditions, specifically fire watches, are a hazard is without merit. I I i l i l l j

4

    .                                                                                              .q 22                 .

C. Combustibility The Petitioners alleged that, contrary to NRC regulations. Thermo-Lag is i combustible. The NRC staff recognizes that Thermo-Lag is combustible. To assess Thermo-Lag combustibility, the NRC staff conducted a testing program at the l National Institute of Standards and Technology (NIST) based on the AmerJear! ! Society for Testing and Naterials (ASTM) Standard E-136. Under this testing

,      standard, the material is considered to be " combustible" if three out of four 3

samples tested exceed the following criteria: (1) the recorded temperature of J the specimen's surface and interior thermocouples, during the test, rises

54 'F (30 *C) above the initial furnace temperature; (2) there is flaming from j the specimen after the first 30 seconds of irradiance; and (3) the weight loss of the specimen, due to combustion during the testing, exceeds 50 percent. Of the four Thermo-Lag ~ specimens tested, all experienced a weight loss of greater d

                                                                      ~

than 50 percent and flaming continued in excess of 30 seconds. IN 92-82, . which provided licensees with the results of the E-136 tests and confirmed the

combustibility of Thermo-Lag, restated the NRC fire protection requirements of Section III.G. of Appendix R to 10 CFR Part 50 and asked that licensees review the information for applicability to their facilities.

The NRC's basic. fire protection regulation for comercial nuclear power plants is Section 50.48 of 10 CFR Part 50 " Fire protection." Section 50.48 1 references General Design Criterion (GDC) 3 of Appendix A to 10 CFR Part 50,

      " Fire protection," Appendix R to 10 CFR Part 50 " Fire Protection Program for

Nuclear Power Facilities Operating Prior to January 1, 1979," and various NRC l

fire protection guidance documents. Specifically, Section 50.48(a) states that each operating nuclear power plant must have a fire protection plan that 4

j ,. c 23 satisfies GOC 3, and Section 50.48(b) states that Appendix R to 10 CFR Part 50 establishes fire protection features required to satisfy GDC 3 with respect to certain generic issues for nuclear power plants licensed to operate prior to January 1, 1979.8 These issues are addressed in Section III.G, " Fire protection of safe shutdown capability,Section III.J, " Emergency lighting," and Section III.0, "011 collection system," of Appendix R. Of these three sections of Appendix R, Section III.G addresses the use of fire barriers to protect one train of systems necessary to achieve and maintain hot shutdown conditions in the event of a fire and, therefore, is the regulation of interest here. Section 50.48(a) notes that fire protection guidance for nuclear power plants is contained in two NRC documents. These are (1) Branch Technical Position (BTP) Auxiliary Power Conversion Systems Branch (APCSB) 9.5-1,

                               " Guidelines for Fire Protection for Nuclear Power Plants," for new plants docketed after July 1, 1976, and (2) Appendix A to BTP APCSB 9.5-1,
                              " Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1,1976." These two NRC documents specify preferred methods for fire protection program design including the use of fire barriers to satisfy Section III.G of Appendix R. Fire barriers that meet the criteria of Section III.G of Appendix R to 10 CFR Part 50 and these NRC guidance documents satisfy GDC 3.      NUREG-0800, " Standard Review Plan," (SRP) Section 9.5-1, " Fire Protection Program," incorporates the guidance of BTP APCSB 9.5-1 and Appendix A to BTP APCSB 9.5-1 and the criteria of Section III.G of Appendix R
                                     'While Appendix R is applicable only to facilities that commenced operation prior to January 1,1979, as discussed earlier in this Director's Decision, facilities commencing operation on or after January I, 1979, are bound to satisfy the criteria of Appendix R through license conditions or i                             licensing commitments.

l l l l

3 24 to 10 CFR Part 50. Therefore, fire barriers that meet the guidelines of SRP Section 9.5-1 also satisfy 10 CFR 50.48 and GDL 3. As stated in 10 CFR 50.48(a), the purpose of the fire protection plan is "to limit fire damage to structures, systems, or components important to safety so that the capability to safely' shut down the plant is ensured." In general, a fire protection plan consists of administrative controls and procedures, personnel for implementing the plan and for fire prevention and manual fire suppression activities, fire detection systems, automatic and manually operated fire suppression systems and equipment, and fire barriers. Section III.G of Appendix R to 10 CFR Part 50 is the only part of the fire protection regulations that addresses the use of fire barriers. It addresses the use of fire barriers to protect one train of systems necessary to achieve and maintain hot shutdown conditions in the event of a fire. Fire barriers are required to have either a 1-hour or 3-hour rating depending on l l the specific requirement. However, Section III.G does not provide acceptance l l criteria for fire barriers, nor does it address the combustibility of fire barrier materials. The criteria are set out in BTP APCSB 9.5-1, Appendix A to 1 BTP APCSB 9.5-1, and SRP Section 9.5-1. These NRC documents do not preclude the use of combustible materials for construction of fire barriers required to have a 1-hour or 3-hour rating. On March 25, 1994, the staff consolidated and clarified in Supplement I to Generic Letter (GL) 86-10, the fire barrier ) l criteria specified in the BTPs and the SRP. This GL supplement provides detailed staff guidelines for assessing the combustibility of fire barrier j materials, but it does not preclude the use of combustible materials for fire l barriers required to satisfy a 1-hour or 3-hour rating. In fact, the fire 1 barrier criteria are appropriately focused on the performance of the fire l 1

aj 25 barrier and its ability to achieve its intended design function, that is, its ability to limit temp 6nture rise within the barrier enclosure and to prevent the passage of fla,me or gasses hot enough to adversely affect the functionality of the safe shutdown components (e.g., cables) enclosed within the fire barrier. Thenno-Lag 330-1 is a sacrificial material. When it is exposed to elevated temperatures, such as those experienced during a fully-developed room fire, it sublimes and transitions from a solid to a vapor. The vapors go through an endothermic decomposition process (pyrolysis) which absorbs heat 1 from the fire. As a result of the pyrolysis, the unreacted Thermo-Lag material is replaced by an insulating char layer which is composed of small interconnecting cells having a large surface area. The char layer re-radiates energy and limits heat transfer through'the Thermo-Lag material. The low thermal conductivity of the char layer provides additional thermal insulation. l Therefore, even though Thermo-Lag is classified as a combustible material when testing in accordance with the guidance of Supplement I to GL 86-10, properly designed, qualified, and installed Thermo-Lag can yield fire barriers with a i 1-hour or 3-hour rating which will protect safe shutdown components from the effects of the fire. Therefore, such barriers can satisfy the requirements of 10 CFR 50.48 and GDC 3. l l  ! To provide reasonable assurance that Thermo-Lag fire barriers installed i in the nuclear power plants can meet their intended function, representative l Thermo-Lag fire barrier assemblies have been subjected to full-scale i qualification-type fire endurance tests conducted in accordance with the l guidance of Supplement I to GL 86-10. This guidance provides standard and uniform test methods and acceptance criteria for assessing the fire-resistive j I i l l

r: 26 capabilities of these barriers. The staff has found the use of Thermo-Lag acceptable as a fire barrier material when it is used in accordance with existing NRC regulations and guidance and where supported by appropriate tests and analyses. However, there are two types of applications where the use of Thermo-Lag material is not appropriate. These are (1) enclosing combustible-materials - (e.g., insulated cables) within Thermo-Lag fire barriers to eliminate. the ' combustible materials as a fire hazard and (2) using Thermo-Lag as radiant ' energy heat shields inside noninerted containments. Section III.C of Appendix R (and the equivalent SRP guidance) specifies three options for protecting redundant trains of systems necessary to achieve l and maintain hot shutdown conditions located within the same fire area outside i of containment. Two of the three options (Sections III.G.2.a and c) rely on ' the use of fire barriers with a 1-hour or 3-hour rating, as discussed above. The third option, Section III.G.2.b, specifies the separation of redundant safe shutdown trains by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards. (A typical example of intervening combustibles is a cable tray loaded with cables, because cable jacket materials are combustible.) Therefore, spacial separation, and not fire barriers, are used to meet Section III.G.2.b. However, to meet this requirement, some licensees have enclosed combustibles that are installed between redundant shutdown trains within a fire barrier. In theory, the fire barrier prevents an exposure fire from igniting the intervening combustible materials and spreading along them from one redundant train to the other. Thus the fire barrier effectively eliminates the intervening combustible as a fire hazard. If the fire barrier itself is noncombustible and the redundant

   .      -      - ~       . -        . - - - - - ~ . - . -                          . - . . . - . _ - -               . _ _       . _ . - _ -.

i - 3 l 21 l

safe shutdown trains are separated by a horizontal distance of more than 20 feet, then the configuration meets Section III.G.2.b of Appendix R. However, if the fire barrier material used to enclose the intervening combustibles is also combustible, such as Therwo-Lag, then the licensee has simply installed one combustible material over another and has not eliminated the intervening j fire hazard. In a limited number of cases, licensees have enclosed j intervening combustibles within Thermo-Lag fire barriers under the incorrect 4

l assumption that the Thermo-Lag fire barrier would eliminate the intervening J l combustibles as a fire hazard. Corrective actions will be required in these j cases. J As an alternative to the three options discussed above, Section l III.G.2.f of Appendix R (and the equivalent SRP guidance) provides a fourth l option for noninerted containments, that is, the separation of redundant 4 l safe shutdown components with noncombustible radiant energy heat shields. j { Thermo-Lag is classified as a combustible material when tested in accordance 1 with the guidance of Supplement 1 to GL 86-10. Therefore, it does not meet l the criteria for radiant energy heat shields. Licensees using Thermo-Lag in l this fashion will also be required to take corrective action. 4

To assure that corrective actions are taken in these cases, the NRC 4

2 staff issued IN 95-27. In that IN, the staff addressed enclosing combustible ~

materials within Therwo-Lag fire barriers in an attempt to eliminate the 1

l combustible materials as a fire hazard and using Thermo-Lag to construct 1 j radiant energy heat shields inside noninerted containments. The staff identified such solutions for reevaluating the use of Thermo-Lag for these i j applications as: (1) reanalyzing post-fire safe shutdown circuits inside ! containment and their separation to determine if the Thermo-Lag radiant energy i i J

l- m 28 shields are needed. (2) replacing Thermo-Lag barriers installed inside the containment with nnncombustible barrier materials, (3) replacing Thermo-Lag barriers used to create combustible-free zones with noncombustible barrier materials (4) rerouting cables or relocating other protected components, or (5) requesting plant-specific exemptions ~where technically justified. One of the Petitioners also asserted that subsection Sa(3) of Section 9.5-1 of the SRP states that fire barrier designs "should utilize only ! non-combustible materials." This section of the SRP does not apply to fire barriers which are used to separate redundant safe shutdown components located l l within a nuclear power plant fire area. Rather, it applies to fire barrier penetration seals, which are typically installed in fire area boundaries. Thermo-Lag 330-1 is not used in such applications. The principal consideration for 1-hour and 3-hour rated fire barriers installed to meet NRC fire protection requirements and guidelines is that they l' can achieve their intended design function. That is, that they can limit 1 temperature rise within the barrier enclosure and prevent the passage of flame l or gasses hot enough to adversely affect the functionality of the safe shutdown components enclosed within the fire barriers. The fact that Thermo-Lag material is combustible does not preclude Thermo-Lag fire barriers I from achieving the intended function of preventing fire damage if the fire barriers are properly designed, qua, fied, and installed. The Petitioners' l l contention that Thermo-Lag material should not be used because it is combustible is without basis. l l e

1 29 D. Annacity Deratina The Petitioners assert that Thermo-tag could contribute to starting a fire instead of protecting from it. They further alleged that faulty ampacity l derating factors could result in the use of inappropriate cables that, if undersized, could overheat and cause its insulation to deteriorate. 1 Ampacity derating is the lowering (derating) of the current-carrying capacity of power cables enclosed in electrical raceways protected with fire barrier materials because of the insulating effect of the fire barrier material. This insulating effect may reduce the ability of the cable insulation to dissipate heat. If not accounted for in the plant design, the increased cable insulation temperature could lead to premature insulation failure. Other factors also affect ampacity derating, including the extent of cable fill in the raceway, cable type, raceway construction, and ambient temperature. The National Electrical Code, Insulated Cable Engin'eers Association (ICEA) publications, and other industry standards provide ampacity derating factors for open air installations. These stancards do not provide derating factors for fire barrier systems. Although a national standard test method is in the process of being developed but has not yet been established, i ampacity derating factors for raceways enclosed with fire barrier material are determined by testing for the specific installation configurations. TSI, the manufacturer of Thermo-Lag, has documented a wide range of ampacity derating factors that were determined by testing, for raceways enclosed within Thenno-Lag fire barrier materials. On October 2, 1986, TSI informed its customers that, while conducting tests in September 1986 at Underwriters Laboratories, Inc. (UL), it found that the ampacity derating factors for Thermo-Lag barriers were greater than previous tests indicated. 1 L_. _ -- - - --

I 30 However, the cable fill and tray configurations were different for each test than those tested previously. In addition, the NRC staff learned that UL l l performed a duplicate cable tray test that resulted in en even higher derating factor. The NRC staff also learned of the determination of other derating I factors during its review of other tests' conducted at Southwest Research i Institute (SwRI).' The NRC special review team concluded that the ampacity derating test results completed at the time of the review, including the UL test results, were indeterminate. This conclusion was based on observed inconsistencies in the derating test results of the various testing laboratories. The special review team found that there was no national consensus test standard (e.g., Institute of Electrical and Electronics Engineers (IEEE) or American National Standards Institute (ANSI)) for conducting these tests, and that some licensees had not adequately reviewed ampacity derating test results to determine the validity of the tests and the applicability of those test i 1

              'The test procedures and test configurations differed among the testing laboratories. Therefore, the results from the different ampacity tests may not be directly comparable to each other.                                                                         l The NRC staff is concerned that the ampacity derating factors, as determined in UL tests for Thermo-Lag barrier designs, are inconsistent with TSI results for similar designs because different times were allowed for the temperature to. stabilize before taking current measurements. Inconsistent stabilization times would call into question the validity of previous TSI results. The NRC also noticed during the review of the Industrial Testing Laboratories (ITL) test reports that ambient temperature and maximum cable temperature were allowed to vary widely for some tests. Therefore, those tests in which the ambient and maximum cable temperatures were not maintained within specified limits may be questionable. Additionally, a licensee discovered a mathematical error for the ampacity derating factor published in an ITL test report. A preliminary assessment of the use of a lower-than-actual ampacity derating factor indicates that higher-than-rated cable temperatures are possible for Themo-Lag installations. Higher-than-rated cable temperatures could accelerate the aging effects experienced by the cable.

c-----w- - -- m - ---#~ - - wm

                                                                                                                                     .:p 1

l l 31 results to their plant design. The special review team recognized that, in hypothetical cases, nonconservative ampacity derating factors could have been  ! Instrumental in the installation of inappropriate cables, which as a result, __ could suffer premature cable jacket and cable insulation failures over a period of time. However, since that time, the NRC staff has determined that in practice the ampacity derating factor resulting from Thermo-Lag insulating l l properties represents only one of many variables used in determining the  ! l l design ampacity for power cable systems and that, as discussed below, ' sufficient margin exists in this area to preclude any immediate safety Concerns. For actual installations, various derating factors are typically applied to the ICEA ampacity values provided for each cable size. In general, the cables typically used in actual installations have higher current-carrying capacity than the ICEA ampacity values.' Also, cables are sized based on full-load current plus a 25 percent margin to account for starting current requirements of the load. Given the short duration of typical equipment starts, this margin is available to compensate for any errors in ampacity derating. Further, use of a cable size larger than normal may be required as a result of voltage drop considerations for long circuit lengths. In typical applications this also provides additional current-carrying capacity. Given these conservatisms inherent in the design ampacity of cable systems and in addition the fact that most power cables required for safe shutdowe are not normally energized, but are typically operated during surveillance testing for short time periods, the likelihood that cables could ignite as a result of

                            ICEA ampacity values include conservatisms to compensate for skin and proximity effects and shield and/or sheath losses which may or may not apply l

in specific situations. 1

A 6 32 Therno-Lag ampacity derating errors has been judged by the NRC staff to be unlikely. In addition, based on these conservatisms and the currently available information on existing plants, ampacity design, and operating history, the NRC staff believes that the ampacity derating issue is not an immediate safety issue but rather is an aging issue to be resolved over the long tern." E. Dvster Creek Failed To Report Test Results on Cc *-astibility to the NRC ' The Petitioners requested that Oyster Creek's license be suspended based on the following: (1) SwRI conducted fire tests on Therno-Lag 330-1 specimens for GPUN, the licensee for Oyster . reek, and reported that all specimens ignited approximately 2 seconds after they were inserted into the furnace and failed specified criteria because of flaming after the first 30 seconds of testing, an outside temperature rise higher than 30 *C, and a weight loss of l 50 percent; (2) GPUN's operation of Oyster Creek with knowledge of the SwRI report is an example of GPUN's reckless disregard for fire protection and public safety; (3) in the event of fire, Thenno-Lag is likely to fail its intended function of protecting vital electrical cables running from the control room to plant safety systems used to shut down the reactor; (4) current installations of Thermo-Lag are likely to fail in less time than the 1 hour (when smoke detectors and automatic sprinkler systems are present) or 3 hours (when there are no fire detection and suppression systems) that NRC regulations require for fire barriers to withstand fire; (5) the NRC Inspector

                       " Generic Letter 92-08 requires licensees to review the ampacity derating factors used for all raceways protected by Therno-Lag 330-1 (for fire protection of safe shutdown capability or to achieve physical independence of electrical systems) and to determine whether the ampacity derating test results relied upon are correct and applicable to the plant design.

Presently, the staff is conducting reviews of followup actions to close out ampacity derating concerns with licensees pursuant to GL 92-08.

  .                                                                                    M 33                -

General issued a report in August 1992 condemning NRC's handling of the Thermo-Lag issue and documenting the NRC staff's failure to understand the ' scope of the problem; (6) in April 1994, ITL and its President pleaded guilty to five felony counts of aiding and abetting the distribution of falsified test data; (7) on September 29, 1994, the U.S. Department of Justice issued a 4 seven-count indictment against the manufacturer of Thermo-Lag and it's* thief-Executive Officer for willful violations of the Atomic Energy Act, conspiracy to conceal material facts, and making false statements to defraud the United States, in connection with $58 million in fire barrier material; (8) GPUN has known since at least August 11, 1992, that Thermo-Lag 330-1 as a structural base material is combustible and that it was 'in violation of Appendices A and R to Part 50 of Title 10 of the Code of Federal Reaulations (10 CFR) and the NRC Standard Review Plan, NUREG-0800; (9) GPUN failed to report the SwRI test - results in response to GL 92-08 of February 10, 1994, when asked to describe t l the Thermo-Lag 330-1 fire barriers installed as required to meet 10 CFR Part 50, Appendix R; and (10) continued reliance on fire watches at Oyster j Creek is an unreasonable and unnecessary hazard to the public health and safety because of an inoperable fire protection system for safe shutdown of

the reactor and installed combustible material on the shutdown systems.

j i Several of the ' issues listed above have been addressed earlier in this decision. Therefore, the NRC staff will only address below the remaining plant-specific issues. As discussed earlier in this decision, the NRC issued IN 92-82 to inform the industry of the results of combustibility tests performed by NIST in early August 1992. These tests confirmed the combustibility of Thermo-Lag. As a result of discussions with the NRC staff 1 on the subject of Thermo-Lag combustibility, GPUN decided to independently i i

d

     ,                                                          34 verify the results of the E-136 tests performed by NIST and contracted SwRI to j               perform the E-136 tests. The results of these tests, as documented by the telecopy transmittal sheet submitted with the Petition, confirmed the j               combustibility of Thenno-Lag.              Contrary to the Petitioners' allegations, the j               NRC staff does not require that licensees report the results of their independent testing.      It should be noted here that, prior to the SwRI testing j              that confirmed combustibility, the NRC was aware of the combustibility of Thermo-Lag ano that the NRC was also well aware of the results of the E-136 i                                                                                                                 l tests performed by GPUN through telephone conversations with GPUN personnel, even though there was no requirement for GPUN to report these test results.

i 4 The Petitioners also alleged that GPUN did not report to NRC its i findings of the swr 1 test results in its " Response to Request for Additional Information Regarding Generic Letter 92-08, 'Thermo-Lag Fire Barriers,'" (RAI)

dated February 10, 1994.

The RAI quoted by the Petitioners did not request that GPUN report to NRC its findings of the swr 1 test results and, in addition, the NRC staff does not require that licensees report the results of their independent testing. Therefore the NRC staff has concluded that, contrary to the Petitioners' allegation, GPUN did not have to report to the NRC its findings of the SwRI ! test results. For the reasons stated above, the suspension of Oyster Creek's license, i as requested by the Petitioners, is not warranted.

!            F. Dry-Joint Issue at Comanche Peak Unit 1 The Petitioners requested that (a) the Comanche Peak Unit I license be suspended, (b) the licensee perform additional destructive analysis for Thermo-Lag configurations, and, (c) the licensee perform fire tests on 4

1

1 i.

      -                                                                   35 l                              upgraded " dry-joint" Thermo-Lag configurations based on the following:

(1) the licensee's records on the original installation of Thermo-Lag fire i barriers on conduits and cable trays indicate that its contractor followed i specifications for pre-buttering all joints; (2) NRC Inspection Report Nos. ! 50-445/93-42; 50-446/93-42 found, based on destructive analysis documents, that a concern did exist where Thermo-Lag conduit joints fell apart easily and did not appear to have any residual material of a buttered surface, indicative of a joint that had not been pre-buttered; (3) the " dry joint" deficiency appeared in Room 115A and other areas of the unit; (4) the licensee directly contradicts an NRC inspector's findings that were determined in part by destructfve analysis; (5) the " dry joint" or absence of pre-buttering of Thermo-Lag panels can be dete: mined only by destructive analysis and cannot be determined by a walk down visual inspection; (6) the findings reported in the Comanche Peak Unit 1 Region IV Inspection Repcrts 50-445/93-42 and 50-446/93-42, based on the limited amount of destructive analysis conducted at the unit, constitute a substantial documentation of installation deficiencies found in Thermo-Lag fire barriers as documented in NRC IN 91-79 and Supplement 1; (7) neither the NRC nor the industry, by its agent NEI, nor a utility, have conducted fire tests on dry fitted or " dry joint" upgraded configurations of Thermo-Lag 330-1; and (8) the presence of " dry joint" upgraded configurations in Cocaanche Peak Unit I constitutes an untested application of Thermo-Lag fire barriers. - These allegations were based on the Petitioners' interpretation of NRC Inspection Report 93-42 issued on February 21, 1994. By letter of November 29, 1994, TU Electric, the licensee for Comanche Peak Unit 1, sent a letter to the NRC staff responding to the Petition. e a

l. i *^ 1 36 ) The tem " joint" refers to the interface between two adjacent Thermo-Lag i surfaces. 1 Comanche Peak Unit 1 installation procedures for Thermo-Lag fire { } barriers specify that, during the initial installation process, the joints j should be pre-buttered (or covered) with Thermo-Lag trowel grade material i

before the mating surfaces are joined to ensure adhesion of the surfaces. The term " dry joint" refers to the lack of Thermo-Lag trowel grade mate d in'a j joint.

The failure to pre-butter a joint with trewel grade Thermo-Lag could result in a weakening of the joint during a potential fire exposure and could i

i provide an exposure path in the fire barrier envelope. The NRC performed an inspection at Comanche Peak Unit 1 on November 2-5, and 23-24, 1993, and j January 26-28, 1994, to compare the Thermo-Lag test specimens with the j upgraded Thermo-Lag configurations on site. The results of this inspection are documented in NRC Inspection Report 93-42. The report stated that there ' i appeared to be a large nui.$er of deficiencies with the installed fire barriers i and that an example of these deficiencies involved dry joints on conduit I g overlays installed on pedestal hangers. The NRC inspector did not personally observe the dry joints in question. His statements were based on observations l j made by TU Electric and documented in an Operations Notification and 1 j Evaluation (ONE) form. However, the ONE form in question did not identify a dry joint. Instead, the ONE form identified a condition that was - i i conservatively reported as an apparent dry joint. Upon further evaluation of l I the ONE form, TU Electric detemined that the joint in question had in fact l j been pre-buttered with trowel grade Themo-Lag. These facts are discussed in 4

more detail below.

i

!                               On November 25, 1992, a speed memo was written by a TV Electric j                contractor identifying " apparent unsatisfactorily conditions on Unit 1 i

i 4 1 i )

       . , . . . . . _ - .                      _ . . , _   ,,       _                             _ _ _ _ _ . , - _ _ _ . . . _ _ . _ _ . _ _ _ , _ _ _ _ _ _ - - . , . _ _ _ ~ _ _ _ _ . _

i j . i

• 37 I

comodities. " This memorandum identified "an apparent" dry joint on an j oversize coupling section (on top of a pedestal hanger). The speed memo also ' stated that, "we have decided that the best vehicle to call attention to these i i apparent deficiencies would be a letter to your attention for further l evaluation of the situation...." The letter was forwarded to the appropriate l i TV Electric engineering section. i , j i The cognizant TU Electric engineer performed a walkdown of the described j areas and evaluated the commodities. He conservatively initiated a ONE form l i (the process used by TU Electric to report problems and develop resolution for , t j the identified problems). A comprehensive evaluation of this condition determined that the joint had been pre-buttered. Therefore, the engineering resolution for this condition was that "this is not a deficient condition, and there are no generic implications." i The originator of the speed memo initially believed that the condition in question was a dry joint because of the appearance of the joint. During alignment of Thermo-Lag panels, the leading edge of one panel contacts the outer edge of a preceding panel and forces most of the trowel grade along the initial contact edge toward the inside of the Thermo-Lag envelope. Subsequent shrinkage of the trowel grade in the joint can give the appearance of a dry joint because the trowel grade material is not visible. Therefore, contrary to the Petitioners' allegation, there was no " dry joint" deficiency on the pedestal hanger. The Petitioners also alleged that dry joints appear in other Thermo-Lag installations at Comanche Peak Unit 1. In response to the Petition, TU Electric performed an electronic search of its ONE form data base. The search did identify additional ONE forms related to dry joints. However, Thermo-Lag

g 38 rework crews and the quality control inspectors at Comanche Peak Unit I have used the term " dry joints" and "no visible trowel grade material" synonymously. Upon further investigation of these ONE forms, it was determined that trowel grade material had in fact been applied to the joints in question. Therefore, these ONE forms were also dispositioned as "not a nonconforming condition." These findings support the NRC staff's conclusion 1 i i that, contrary to the Petitioners' allegations, there is no evidence of dry ' joints at Comanche Peak Unit 1. The Petitioners' allegations regarding dry joints at Comanche Peak Unit I are based on premises that are faulty and contrary to the information contained in Inspection Report 93-42. In regard to the Petitioners' request that the licensee perform fire tests on upgraded " dry joint" Thermo-Lag configurations and additional destructive analysis, the NRC staff has reviewed the documentation provided by the licensee in response to the RAls regarding GL 92-08 and concluded that the l licensee's quality assurance program gave adequate confidence that the as-installed Thermo-Lag configurations at Comanche Peak Unit I conform with NRC specification requirements for both material and installation attributes. Accordingly, suspension of the Comanche Peak Unit I license, as requested by the Petitioners, is not warranted. G. Protection of Rubin Feldman The Petitioners assert that, rather than protecting the public, the NRC is protecting Rubin Feldman, President of the company that manufactures Thermo-Lag. As discussed earlier, the NRC received aliegations in 1991 that questioned the adequacy of Thermo-Lag fire barriers. In response (1) the Office of ths Inspector General (0!G) and the Office of Investigations (01) r , ,

3 l , 39 fomed a joint task force to investigate the allegations and (2) the Office of Nuclear Reactor Regulation (NRR) established a special team to review the safety issues raised by the allegations. Throughout its review, the special team gave expert technical advice and assistance to the OIG/01 task force. The Director of NRR tasked the NRR staff to resolve the technical issues l raised by the special team. The NRC staff continued to cooperate fully with the investigative task force. Further, the NRR staff carried out a full-scale test program and developed other technical data and information for the investigative task force. These NRC staff efforts contributed significantly to a referral to the Department of Justice of possible wrongdoing by TSI. The l referral resulted in a seven-count criminal indictment of TSI, the manufacturer and supplier of Thermo-Lag fire barriers and of its President, . Rubin Feldman, by a Federal Grand Jury. The NRC staff continued to support the Department of Justice throughout' the criminal case.it In addition, throughout the trial, the NRC staff continued to pursue corrective actions consistent with its action plan for the resolution of the Thermo-Lag issues. The above facts contradict the Petitioners' assertion that the NRC was protecting Rubin Feldman. ' H. NRC Seemina Comolicity with Utilities The Petitioners also assert that there is seeming complicity between the NRC and the ~1icensees and that licensees seek to avoid costly replacement of the Thermo-l.ag. In May 1991, the NRC Office of the Inspector General performed an inspection of the NRC's staff performance in rcgard to Thermo-Lag barriers and

                            '3 The jury returned a verdict of "not guilty" on all counts of the indictment against TSI and Mr. Feldman.

l

i . j is

40 i

found indications of inadequate performance by the NRC staff in the acceptance 1

and review of Thermo-Lag barriers. Subsequently, the NRC staff initiated an aggressive program of corrective actions to rectify the deficiencies identified in the review and response process, as summarized earlier in this decision.

In addition, the staff has expended considerable time and effort to j address and resolve Thermo-Lag issues to ensure that licensees return to compliance with existing NRC fire protection requirements. The NRC staff issued three requests for additional information regarding GL 92-08 to each licensee using Thermo-Lag to obtain information on the specific Themo-Lag

material installed at each plant, details about the wrective actions each licensee intended to take to return to compliance with HRC fire protection j requirements, and schedules for the implementstion of these corrective j actions. The response of each licensee was evaluated by the NRC staff. As a j

consequence of this substantial NRC staff effort, a number of licensees have ! l i already returned to compliance with NRC requirements by a variety of means ! which include replacing, rerouting, or upgrading existing Thermo-Lag barriers, ( I ) performing post-fu e safe shutdown reanalysis, and installing additional fire detection and suppression features. All of these measures involve some burden i on licensees. In addition, some licensees have initiated costly programs to j perform plant-specific fire endurance tests of other fire barriers with the j intention of replacing Thermo-Lag with these barriers. All licensees who 4 utilize Thermo-Lag will need to expend resources consensurate with their , reliance on Thermo-Lag to come into compliance with NRC fire protection i j requirements. NRC staff oversight will ensure that this is the case. 1 i i i l i i l

__ _ __.._ _ _ _ _ _ . ~ _ ___._ . _ _ . . . _ - _ . _ _ _ _ ._ . . _ . _ _ _ _ _ \

     .                                                                                                                             '9
   'J 41 The Petitioners' assertion of seeming complicity with utilities on the t                         part of the NRC staff is unfounded in the light of the significant NRC staff efforts to ensure that licensees expend the resources necessary to return to                                  i i

compliance with NRC requirements. I { IV. CONC'USION L The Petitioners request that the NRC order the immediate shutdown of all reactors using Thermo-Lag and the suspension of Oyster Creek, Peach Botton l Units 1 and 2, and Comanche Peak Unit 1 operating licenses. For the reasons discussed above, I find no basis for taking such actions. 1 Rather, on the basis of the review efforts by the NRC staff, I ( conclude that the issues raised by the Petitioners are being addressed by l licensees in a manner which assures adequate protection of the public health h and safety. Accordingly., the Petitioners' requests for action pursuant to 10 CFR 2.206 are denied. A copy of this Decision will be placed in the Commission's Public ) Document Room, Gelman Building, 2120 L Street, N.W., Washington, D.C., and at the Local Public Document Room for the named facilities. A copy of this Decision will also be filed with the Secretary for the Commission's review as provided in 10 CFR 2.206(c) of the Commission's regulations. l

   ,  . _ _ _ . ._-        .. -    -     _ _ - -   _ - . . - _       - . = . - _ _

i k v I 'J !. 42 ) As provided by this regulation, the Decision will constitute the final action of the Commission 25 days after issuance, unless the Commission, on its

;                   own motion, institutes a review of the Decision within that time.

Dated at Rockville, Maryland this 3rd day of April . 1996. I, FOR'THE NUCLEAR REGULATORY COMISSION i h0 William T. Russell, Director ! Office of Nuclear Reactor Regulation i i i i l i 4 1 ',i i L._ _ _

qi

 's 7590-01 UNITED STATES NUCLEAR REGULATORY COMIS11Qg ALL LICENSEES OF REACTORS WITH INSTALLED IliEEMO-LAG FIRE BARRIER MATERIAL ISSUANCE OF DIRECTOR'S DECISION UNDER 10 CFR 2.206 Notice is hereby given that the Director, Office of Nuclear Reactor i

Regulation, has acted on Petitions for action under 10 CFR 2.206 received by a

letter dated September 26, 1994, from the Citizens for Fair Utility Regulation and the Nuclear Information and Resource Service; by a press release dated October 6,1994, from the Maryland Safe Energy Coalition; by separate letters l dated October 21, 1994, from the GE Stockholders' Alliance and Dr. D. K.

Cinquemani; by a letter

• dated October 25, 1994, from the Toledo Coalition for Safe Energy; by a letter dated October 26, 1994, from R. Benjan; by a letter dated November 14, 1994, from B. DeBolt; and by a letter dated December 8, l 1994, from the Nuclear Information and Resource Service and the Oyster Creek i Nuclear Watch. The Petitioners requested that the U.S. Nuclear Regulatory Commission (NRC) take action with regard to the use of Thermo-Lag by reactor licensees and that their letters be treated as Petitions pursuant to Section 2.206 of Title 10 of the .Gpde of Federal Reaulatigni (10 CFR 2.206).

The Citizens for Fair Utility Regulation and the Nuclear Information and Rasource Service requested (1) Texas Utilities Electric Company, the licensee of Comanche Peak Steam Electric Station, Unit 1, perform additional destructive analysis for Thermo-Lag configurations in proportion to the total installed amount to determine the degree of " dry joint" occurrence, (?) the licensee perform fire tests on upgraded " dry joint" Thermo-Lag configurations for conduit and cable trays to rate the barrier as a tested configuration in

4

       .                                                                                                                 1 1

i a 2 I compliance with fire protection regulations, and (3) the NRC immediately ' suspend the Comanche Peak Unit I license until the above listed corrective

                   ~ actions are taken.

2 The Naryland Safe Energy Coalition requested imediate l l ) shutdown of both reactors at the Peach Bottom plant until the risk of fire j near electrical control cables due to combustible insulation is corrected. i i Dr. Cinquemani and the Toledo Coalition for Safe Energy requested that the NRC immediately shut down all reactors where Thermo-Lag is used until it has been j removed and replaced. The GE Stockholders' Alliance requested shutdown of all l 1 reactors where Thermo-Lag is used until it has been removed and replaced with fire-retardant material meeting NRC standards. R. Benjan requested imediate I shutdown of all reactors where Thermo-Lag is used. B. DeBolt requested shutdown of all reactors in which Thermo-Lag is used until it has been removed ! and replaced. The Nuclear Information and Resource Service and the Oyster 1 j Creek Nuclear Watch requested that NRC imediately suspend GPU Nuclear Corporation's (GPUN's) operating license for Oyster Creek Nuclear Generating Station (OCNGS) until GPUN removes Thermo-Lag fire barrier material and replaces it with a competitive product that meets current NRC fire protection regulations. The Director of the Office of Nuclear Reactor Regulation has determined that these requests should be denied for the reasons stated in the " Director's Decision Under 10 CFR 2.206" (DD-96-03), the complete text of which follows this notice, and which is available for public inspection at the Comission's Public Document Room, the Gelman Building, 2120 L Street, N.W., Washington, D.C., and at the Local Public Document Room for the named facilities. A copy of this Decision has been filed with the Secretary of the Comission for the Comission's review in accordance with 10 CFR 2.206(c) of

         .,--~y..
          -                  -         .- - - , -.         -        g  ,, _         ,,                                m     -

                                                                                    'fg' d

o 3 the Comission's regulations. As provided by this regulation, this Decision will constitute the. final action of the commission 25 days after the date of l issuance unless the Commission, on its own motion, institutes review of the m Decision within that time. Dated at Rockville, Maryland, this' 3rd day of April , 1996. FOR THE NUCLEAR REGULATORY ComISSION b/ William T. Russell, Director Office of Nuclear Reactor Regulation

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1 l Maryland Safe Energy Coalition P.O. Box 33111 i Baltimore, MD 21218 4 410 243 2077 i FAX 235-5325 j ro0001@epfl2.epfibalto.org k C* PA M & cx.6 P Y py pg yyp i 1 N OC 2 PACFS Fo Lu*wM

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j- .. - 1 i; Maryland Safe Encrgy Coalition j P.O. box 33111 4 Bakimore, MD 21218 410 243 2077 FAX 235 5325 i; ro0001@cp02.epubal(0.org 1 k) t PRESS RELEASE j october 6, 1994 THE MARYLAND SAFE ENERGY COALITION CALLS FOR THE IMMEDIATE SHUT DOWN STATION OF IN BOTH DELTA,NUCLEAR REACTORS AT THE PEACH BOTTOM ATOMIC POWER PENNSYLVANIA. } THE RISK OF A FIRE NEAR THE i ELECTRICAL CONTROL CABLES DUE TO COMBUGTIBLc., INSULATION, WHICH COULD CAUSE A CATASTROPHIC HELT DOWN, IS UNACCEPTABLE. AS

}

REPORTED ON THE FRONT PAGE OF THE BALTIMORE BUN. (SEPT. 30), THIS j SAFETY VIOLATION WAS 80 SERIOUS THAT THE MANUFACTURER OF THE 1 FLAME RETARDANT WAS INDICTED ON CRIMINAL CHARGES. i NHILE THIS HAZARD EXIST 8 IN AT LkAST 60 OTHER U.S. REACTORS, THE f SAFETY OTHER CONDITION SERICUS AT PEACH BOTTOM IS PURTHER HEIGHTENED BY SEVERA VIOLATIONS. WE DEMAND THAT THE REACTORS AT PEACH { BOTTOM ARE CORRECTED. BE REPT SKUT DOWN UNTIL THE ABOVE AND FOLLOWING CONDITIONS t j i IN ADDITION TO THE ABOVE HAZARD, CRACKS WERE DISCOVERED IN THE { STRUCTURAL 3, SUPPORT (CORE SHROUD) OF THE REACTOR FUEL IN REACTOR

• INDICATING VESSEL. WE SUPPORT POSSIBLE CRACK 8 IN OTHER PARTS OF THE REACTOR THE RESOURCE SERVICE (N!8ts), DEMAND OF THE HUCLEAR INFORMATION

! THAT ALL 8AFETY CLASS COMPONENT PARTS IN BOTH REACTOR VESSELS BE INSPECTED. THIS INCLUDE 8 THE C00LINC l SYSTEM, HEAT TRANSFER SYSTEM AND THE CORE ITSELF, WHERE SHIFTING i PARTS COULD CAUSE THE RODS TO STICK (16 PARTS IN EACH REACTOR) . WE ALSO SUPPORT THE DEMAND BY NIRS THAT THE PEACH BOTTOM LICENSE j BE SUSPENDED UNTIL AN ANALYSIS OF THE 8YNERGISTIC EFFECTS OF CRACKS IN MULTIPLE PARTS IS CONDUCTED. . i IN ADDITION, THE NUCLEAR REGULATORY COMMIS8 ION (NRC) DISCOVERED l THAT BOTH HOUR ON AUGUSTREACTORS HAD NO EMERGENCY COOLING WATER FOR ABOUT AN 3, 1994, WHICH MEANS THAT THE REACTORS COULD HAVE j MELTED DOWN IF THEY OVERHEATED. ! ACCORDING To AN NRC REPORT (AUG. 16, 1994), OTHER CHRONIC I PROBLEMS AT PEACH BOTTOM INCLUDEr C00LINC TOWER LEAKS, COOLANT

  ]                 INJECTION SYSTEM VIBRATION, INJECTION VALVE FAILURES, FEEDWATER i

VIBRATIONS AND LEAKAGE, FUEL POOL HOT SPOTS, INCORE PROBE FAILURES, AUXILIARY BOILER UNRELIABILITY, VALVE FAILURES, AIR 1 BOLEN0ZD FAILURE, AND HYDRAULIC,LEARS AND MALFUNCTIONS. i

• Contact MSEC (above) or NIRS at 1424 16th St.,N.W., Suite 601, j Washington, DC 20036. Phone 202-328-0002.

i l

._. - . - - - - --. .- . . - - - - . - . . ~ _ . . . _ . . . . . . . _ - - ... _ _..~....__...,__,. _ , lilEllIER'is I l0Bll!ll1

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Docket Nos. 50-277/278 N DIS 11 RUT 10N El. Jordan Docket Fih BGrimes NRC PDR JPartlow l.ocal PDR ACRS (10) PO*P Rending OEL D l Mr. Edward G. Bauer, Jr. RBernero Gray file l Vice President and General Counsel SNorris GJohnsnn Philadelphia Electric Con <pany GGears SI ee 2301 Market Street 4 Philadelphia, Pannsylvania 19101 1 Ocar Mr Bauer: Re Peac h Bo t tom A torr.i c Power S t a t i nn , Units . and j 10 CFR 50.55a(g) requires that applicable ASME Code components of a hniling a water-cooled nuclear power facility meet the cr'ouirements set forth in Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. Of the ASME Boiler and Pressure Vessel Code. Each facility is required to have an inspection program plan which is updated every 10 years to meet the requirements of the latest approved applicable edition and i addenda of Section XI. i By letters dated June 28, 1984, December ti, 1984, January 31, 1985, February 1, 1985 ano March 6, 1965, the Philadelphia Electric Company (PECe) submitted to the NRC the Inservice inspection (151) Plan and additional information for the second 10-year interval (Jnly 6, l'/84 to July 6,1994 for tJnit 2 and December 13,19U. tn December 13, 1994 f or tini t J) f or P ach Bot ton: Atomic Power Station, units 2 and 3. The above cit"d sut>mittals also requested relief for Peach Bot tom Aton if Pnwer S l et t i o ri ,

   !hiits ? and 3, f rom certain of the Set.t ion Xl Code requirements.

As indicatftd in the enclosed Safety fvaluation (5L) and the supporting Technical Ivaluation Report (TER) prepared by our contractor, Science Applications International Corporation (SAiC), we have concluded that relief should be granted from the examination requirements which we have determined to be impractical to perform and, by granting relief, would nnt compromise the safety of the facility. The entlosed SE and T[R provide the details of our review, and Attachment 1 to the ',L provides a sunnary tabuldtion of our conclusions concerning o.tch rer:uest. We have de t. ermined pursuant to 10 CFR Part ',0.55a(<;)((,)(i), t hat t hr- i t ene. for which relief is requested are improctical and that the granting of thi, relief i', ,tuthnrited by Idw arid will fint endarnpar life ne property nr the 8 08 DOCK 05000277 n POR

 ,    ,                                                                                Station Support D$gartm:nt             l l

l  % N

                                                                                                                              \

l l IECO ENERGY = a,T r = ;L e,, 965 Chesterbrook Boulevard ! Wayne. PA 19087 5691 ! March 14,1994 Docket Nos. 50-277 50-278 Uconse Nos. DPR-44 DPR-56 1 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Subject:

Peach Bottom Atomic Power Station, Unit 3 Evaluation of Core Shroud Indications

Reference:

(1) Letter from G. A. Hunger, Jr. (PECO Energy) to NRC dated October 29,1993 (2) NRC Meeting Summary of November 3,1993 Meeting Issued December 2,1993

Dear Sir:

In the above referenced letter, PECO Energy Compc7y submitted preliminary information regarding the evaluation of the Indications identifled on the Peach Bottom Atomic Power Station (PBAPS), Unit 3 core shroud. This issue was discussed with the NRC at a meeting on November 3,1993. At the meeting, questions were raised conceming the applicability of the quadrant screening criterion. PECO Energy agreed to revise the evaluation to clarify the significance of the quadrant screening criterion. This clarificattoriis provided in the enclosed GENE-523141 1093, Rev.1 report entitled, " Evaluation and Screening Criteria for the Peach Bottom Unit 3 Shroud Indications.' Additionally, the basis for calculation of shroud crack growth ) rates at PBAPS has been included in the revised report. j If you have any questions, please do not hesitate to contact us.  ; I Very truly yours,

               .        .                  i.

G. A. Hunger, Jr., Director  ! l Ucensing Section MCK/eas l Enclosure cc: T. T. Martin, Administrator, Region I, USNRC W. L Schmidt, USNR esident inspector, PBAPS I I i 940 0103 940314 ADOCK 05000278 q , PDR i - \ / > l

 ,    a I

o ' e GENE-523-141-1093, Rev.1 DRF 137-0010-6 l l Evaluation and Screening Criteria for the Peach Bottom Unit-3 Shroud indications December 3,1993 l Prepared by: " /- /# ~ - T%cos L. Herrera, Principal Engineer Structural Mechanics Projects j

                                                                      ~

l Dr. Hardayal Mehta, Principal Engineer ) Structural Mechanics Projects ) Approved By: M Dr. Sampath Ib(nganath, Manager Structural Mechanics Projects GE Nuclear Energy San Jose, CA 940323 940314 PDR OCK 05000278 PDR

C IMPORTANTNOTICEREGARDING CONTENTS OF TIIIS REPORT Please Read Carefully The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between Philadelphia Electric Co. and GE, and nothing contained in this document shall be construed as changing the , contract. The use of this information by anyone other than PECo, orfor any purpose other than thatfor which it is intended under such contract is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, and assumes j no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights. I l O

 .     .- .                                -         _ .-.- ---                             - - . -                    - - - - . _ _ = _ -                                                                      _- .             .-

t l l t i j Table of Contents l l t 1 1 1 t l EXECUTIVE

SUMMARY

iv i l i I l 1. 0 INTRO D U CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 I l ' \ \ 1.1 B a ckgrou nd . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 l 1.2 S creening Crit eria. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . .. .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . 3 i 1. 3 Re fe r e n c es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

2. 0 FAB RIC ATI ON HI STORY . . . . . . . . . . . . . .'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .

3.0 CHEMISTRY AND FLUENCE CONSIDERATIONS.... . . ................................ 23 l 3.1 Water C hemistry History . . . . .. . . . . . . . . . ... . . . . . .. .. ....... . .. . . . . . . . . . . .. . . .......... . .. . . .. . . 23 ! 3.2 Fluence Consid erations.. .... ........................................ ..... .. ..............25 l 3 . 3 R e fe r en c e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 l l l 4.0 IN-VES SEL VISUAL INSPECTION.. ....... ....... ......... ..... .............. ................ ... 3 6 I 5 . 0 FL AW EVALU ATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . .. . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 9 5.1 S t ru ctu ral Analysis . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . ... . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . 3 9 5.2 Allowable Through-Wall Flaws .......................................... ................. . .. 43

5. 3 S creening C rit eria. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 4 6 5.4 Summary of Screening Criteria............ ..... ........................ . ....r... ..... . . 48 5.5 Application of Screening Criteria................................... ... ..................... . 49 5 . 6 Re fe r e nc es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 0 6.0

SUMMARY

AND CONCLUSIONS .. . . ... ............... ........ .... .... .. .. ....... . .... ... 5 7 i l APPENDIX A DETERMINATION OF THE EFFECTIVE FLAW LENGTH APPENDIX B BASIS FOR CRACK GROWTH RATE l i p 4 l 1 l l l

i

    ~

GENxclear Energy GENE.323141 1M3, Rev.1 EXECUTIVE

SUMMARY

Indications have been observed in the Peach Bottom Unit 3 core shroud. Indications were seen during in-vessel visual inspection (IVVI) of the various shroud welds as recommended by GE SIL 572, Rev.1. Results showed that both circumferential and axial , indications were present at the H3 and H4 welds. H3 corresponds to the weld between I the top guide support ring and core shroud cylinder, and the H4 weld is located at approximately the mid-height of the fuel. In addition, circumferential indications were l observed in the shroud plate associated with a vertical weld. The lengths of the indications associated with the vertical welds were short (s2.5" max.) compared to those l associated with the horizontal welds. ( This evaluation was performed to disposition the indications by demonstrating that the structural integrity of the shroud is maintained for the next fuel cycle (two year cycle with ) power rerate conditions). In addition, the report documents material, water chemistry and j l l fluence information which are additional variables which may have contributed to the i ! shroud condition. ) l The primary focus of this report is to demonstrate that even with several conservatisms in the evaluation, the structural integrity of the shroud is maintained during a limiting event. This was performed by developing conservative screening criteria, assuming throughwall l indications, which can determine the acceptability of the flaws based solely on the IVVI

results. The assumption of through-wall indications removes any uncertainty regarding sizing and the need to further characterize the indications. By meeting the screening criteria, the ASME Code Section XI safety margins are satisfied.

The screening criteria use both linear elastic fracture mechanics (LEFM) and limit load concepts to determine acceptable through-wallindication lengths. The limiting flaw length based on either LEFM or limit load was used for the screening criteria. l The screening criteria also use the ASME Code Section XI criteria for combining flaws based on the proximity ofindications. In addition, a second method for including the f . l interaction between neighboring indication 11ps was considered for the LEFM allowable j flaw size cajeulation. The resulting effective flaw lengths were compared against the screening criteria to determine if the structural integrity of the shroud was maintained. l l iv 1 .

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• l l

GENxclear Ezery GENE.533.H].1093, Rev. ! I l l . I Based on the results of the application of the screening criteria to the observed indications, it is concluded that the structural integrity of the shroud is maintained for the next fuel cycle. All effective indication lengths were shown to be less than the allowable flaw size. l l ! I l l i I i I l b l f 1 V

__. . ._ _.._ __ - __- _.-_ _ _._._ _ _ - - . -_.. ~. _.. _ . . _ __._, m . - _ . _ GENuclear Energy GENE-383-141-1M3, Rev.1 7

1.0 INTRODUCTION

l l The objective of this report is to document the conditions found on the Peach Bottom Unit-3 shroud, and evaluate these conditions based on GE SIL 572, Rev.1 (Reference 1-

1) recommendations, in order to validate the structural margins of the shroud.

Recently, in-vessel visual inspection (IVVI) of the Peach Bottom Unit-3 shroud revealed indications in the inside surface heat affected zones (HAZ) at weld locations H3 and H4. Figure 1-1 is a schematic illustrating the general locations of the shroud welds in Unit-3. Figures 1-2 and 1-3 are the shroud maps which show the locations where indications were found. Figure 1-4 is a plan view which indicates the locations referred to in Figures 1-2 and 1-3. Figure 1-2 shows the inside surface shroud map and Figure 1-3 shows the outside surface shroud map. Horizontal and verticalindications were seen associated with the H3 and H4 welds. Circumferentialindications were also observed associated with one I of the vertical welds (V3). However, these indications were relatively short compared to those associated with the horizontal welos.  ! l l In addition to the H3 and H4 welds, IVVI of the H1, H2, HS, H6, H7 and H8 welds was l performed on the outside surface. Only a few short indications were observed on the l outside surface of H1 and H4. It should also be noted that the area adjacent to the H9 . weld was visually inspected as part of the access hole cover inspection (AHC) at this outage. The inspection did not reveal any indications. Additional detail of the IVVI results is presented in Section 4.0. . l l GE SIL 572, Rev.1, provides the following recommendations based on the observed indications and evaluations performed to date: Plant Fabrication and Ooerational History Review plants fabrication and operational histories for the core shroud, including the materials of construction. Non-Destructive Examination Actions Visual examinations of accessible areas should be performed on the shroud ID and OD l ! surface at the next scheduled refueling outage for all plants with Type 304 stainless steel ! shrouds with six or more years of power operation, and for all plants with L-grade 1

i . . l i

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GENuclear Energy GENE-583-H1-1093, Reu 1 stainless steel shrouds with eight or more years of power operation. These examinations should be performed with an enhanced VT-1 system or a qualified UT examination from the outer surface. Ifindications are not observed, examination should be performed at every second refueling outage. Ifindications are observed, the shroud should be examinea and lengths measured during each refueling outage. The SIL also provides a recommended examination process. l Destructive Testing ! A boat or core sample may be necessary depending on the results of the examination. Structural Marcin Analysis Perform a structural margin analysis using the results from the NDE, and, if performed, the destructive analysis. If numerous indications are observed, the need for corrective l action can be assessed using cumulative flaw length structural margin criteria. i Corrective Action Based on the results of the structural margin evaluation, determine if continued operation  ; l is justified for another cycle without repair. If cracking is found and sufficient structural margin remains, examine the shroud during each subsequent refueling outage. i

                                                                                                                                    ?

l This report provides the pertinent information required to demons'trate that continued l I operation of Peach Bottom Unit-3 is justified based on the SIL recommendations noted above. Specifically, the report presents the following information:

                . Fabrication history of the shroud.
                . Water Chemistry and Fluence Considerations
                . In-Vessel VisualInspection
                . StmeturalMargin Analysis
                . Screening Criteria for Application to IVVI results.                                                       .

It is noted that the loads used in this evaluation correspond to those for power rerate. A i two-year operating cycle was used in the determination of crack growth. 2 l l I ,

d GENuclear Doorgy GENE 523-H1-1M3, Rev.1 i i

1.1 Background

j Indications have been observed in the shrouds of three plants to date (including Peach  !

;                      Bottom Unit-3). Cracking was observed in a BWR/4 located outside the United States in

1990. The cracking was confined to the heat affected zone (HAZ) of a circumferential 4

i weld.

l

{ In 1993, the second occurrence of cracking in a shroud was reported. Cracking was l l observed on the inside surface (ID) of the top guide support ring near the H3 weld. The i cracking was approximately 360* around the circumference, in the weld heat affected zone (HAZ), in a material with carbon content of 0.06%. The fluence was estimated as i 1.8x1020 nyt (E>lMev). f 1 !. In addition to the H3 weld HAZ cracking, indications were visually observed at the H1, H2, H4, HS (shroud cylinder) and, H6a weld HAZ (at core plate support ring).  ! f j Indications were seen mostly on the inner surface at H3, H4 and H5. Indications were  ! I seen on the outer surface at the H1, H2 and H6 welds. l 4 l

1.2 Screening Criteria j I

IVVI provides the length characterization of any present indications. Given that non-l 2 destmetive examination (NDE) of every visually detected indication could be difficult and i time consuming, a method of screening indications for subsequent evaluation is required. $ This report presents such a screening criterion. 4 1 The guiding parameter used for the selection of the indications for further evaluation is the

allowable through-wall flaw size, which already includes the safety factors. If all of the visually detected indications are assumed to be through-wall, then the longest flaws, or combination of flaws, would have the limiting margin against the allow able through-wall a

! flaw size. In reality, the indications are likely not through-wall, and Gerefore, the criteria and methods presented in this report are conservative. The result of this procedure will be the determination of the effective flaw lengths which , will be used to compare against the allowa61e flaw size and selection ofindications for more detailed evaluation. The determination of effective flaw length is based on ASME i Code,'Section XI, Subarticle IWA-3300 (1986 Edition) proximity criteria. These criteria t

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l GENuclear Energy GEhT.-523-141-1093, Rev.1 provide the basis for the combination of neighboring indications depending on various geometric dimensions. Crack growth over a subsequent two year operating and power rerate cycle is factored into the criteria. This is conservative since power rerate will not be in effect during the next fuel cycle. l The proximity rules described here also conservatively assume that there is interaction between two perpendicular flaws. It is assumed that circumferential and axial indications could increase the effective flaw length depending on the unflawed distance between them. This effective circumferential flaw length must be compared against the allowable circumferential flaw length. The axial flaw would be compared against the allowable axial flaw length. 1 Flaws are considered in the same plane if the perpendicular distance between the planes is 4" or less. Any flaws which lie at an angle to the horizontal plane should be separated into a circumferential and axial component. These components can then be used separately in the determination of effective flaw lengths. The selection ofindications for further investigation can be performed by evaluating the resulting effective flaw lengths. Indications with effective flaw lengths greater than the allowable flaw sizes would require further characterization by NDE or more detailed analysis. The procedure described here is conservative since all of the indications are assumed through-wall and are being compared against the allowable through-wall flaw size. The report covers the limiting stresses for all the shroud welds (H1 through H8 welds). Therefore, the screening criteria developed here cover all shroud weld indications. A list of conservatisms used in this evaluation is summarized in Table 1-1. a8 l 4 1

GENuclear Euray GENE-SS3-141 1M3, Rev.1 Table 1-1 Conservatisms Included In Screening Evaluation

1. All surface indications were assumed to be through-wall for analysis.

l

2. The screening criteria limit one-fourth of allowable circumferential flaws to any i

arbitrary 90* sector.

3. All indications are assumed to be grouped together for the limit load calculation and no credit is taken for the spacing between indications.

4. ASME Code primary pressure boundary safety margins were applied even though the shroud is not a primary pressure boundary.

5. ASME Code, Section XI proximity rules were applied.

6. An additional proximity rule which accounts for fracture mechanics interaction between adjacent flaws was used (See Appendix A).

7. The highest stress computed for any single location was used for alllocations.

8. Both LEFM and limit load analysis were applied, even though LEFM underestimates allowable flaw size for austenitic materials and is not required per ASME Code Section XI procedures.

9. Fracture toughness measured for similar materials having a higher fluence was used.

10. The bounding crack growth estimated for the next fuel cycle was included in flaw lengths used for evaluation (See Appendix B). .

I1. A proximity rule to account for perpendicular flaws was applied, although not required by Section XI. k

12. Power rerste conditions were used although it will not be in effect during the next fuel cycle.

O e

• 5

I , GENuclear Energy GENE.5231411093, Rev.1 i i l 1.3 References l l ! 1. GE Services Information Letter (SIL) 572, Rev.1, October,1993 t l l l l l l 9 i t 6

l GENE.5231411093, Rev.1 GENucleu Enero Shroud Head Flange H1

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         .        GENuclear Energy                                                        GENE-383o1411M3, Rev.1 i

I l = l 2.0 FABRICATION HISTORY l i l This section describes the fabrication history of the Peach Bottom Unit-3 shroud. Of key interest is the material composition and any activities which could have possibly l contributed to the increase ofintergranular stress corrosion cracking (IGSCC) l susceptibility. Quality assurance records received from the vessel vendor (Rotterdam) were examined in detail to determine the appropriate information. l l Table 2-1 shows the niaterial data for the Unit-3 shroud. The part numbers are identified in the schematic shown as part of Table 2-1. Also shown in the table is the number of pieces for each part, material designation, heat numbers, and carbon content. Figure 2-1 shows the assembly of the shroud. All welds are identified including vertical and horizontal welds. Figure 2-2 through 2-9 show the details of the shroud welds as I labeled in Figure 2-1. The upper, central and lower rings (part numbers 1,3 and 6) are austenitic stainless steel seamless rolled forging. The materialis ASTM A182 - F304. The heat treatment of these rings consisted of heating to 1100*C, holding for 6 hours, followed by water quenching to below 100 C. The carbon content of the rings ranges from 0.03% to 0.035% max. Hardness cleasurements upon completion of solution beat treatment and rough machining of the rings ranged from Brinell Hardness of 137 to 153.

                                                                                                 ~

Each cylinder is made of 2 plate segments formed and welded to drawing requirements. Plate material is austenitic stainless steel made to ASTM A240, Type 304 specifications. The carbon content of the plate material ranges from 0.057% to 0.062% max. The hardness of the plate material ranges from Brinell Hardness of 137 to 155. All welding was performed by submerged arc-welding except H7. The procedure and welder qualification was performed to ASME Section IX requirements. The filler metal met ASTM A-371 Type ER-308 requirements with required carbon content of 0.08% max. The welded joints did not use backing strips but utilized 3 to 4 hand weld passes. , Weld prep surfaces of the base metal were prepared by machining. The backside of the l groove welding was prepared by grinding or gouging followed by liquid penetrant inspection. Final surfaces of the welds were inspected by liquid penetrant examination. 11

GENuclear Ewgy GENE.3331411093, Rev.1 l l l The H7 weld was performed using metalinert gas with Alloy 82 wire. In addition,100% ultrasonic examination of weld H7 was performed. Based on GE Quality Assurance records received from Rotterdam, no abnormal fabrication history was found. General practice during assembly and shipment of the of j the shroud, bracing, temporary welds, and supports are used to help in meeting the joining  ; of the various components and to meet geometric tolerances. Although there is no record i documentation of these practices, it is likely that they were present during fabrication. These actions result in a local effect on material behavior and stress. For example, the I welding of temporary pads would result in a local area of weld residual stress and perhaps I some grinding (cold work). If these local effects contribute to SCC, it is likely that the cracking would be oflesser concern than cracks near the horizontal welds. l l l S e

                                                                                                     .a

• 12

1 Q D2 [ e

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l l 4 Peach Bottom 3 Shroud Data 3 l s Part Part Name quantity Meterlet Heat / Certificate Carbon Comments on l Numiner Number Content - X Material / Process 1 Upper Ring i Place Alet - F304 F1343-ea.tes 0.035 l 2 Upper cylinder 2 Pieces A240 Type 304 3S62-E9987 0.082 3 Central Iting i Piece AtS2 - F304 F1399-SS.587 0.030 - f 4 Central Cylinder 2 Pieces A240 Type 304 5727-Et t4 E057

                             !     ' 5                                                                                                Se37-Et te         0.000 5  Centret Cyttador    2 Piece                                  A240 - Type 304        2170-E212          0.057

{ - aste-Eta 0.Dec l 8 Imwer Ring 1 Place A102-F304 F1400-88.16e 0.03S { l 7 tawer cylinder 2 Ploces A240 Type 304 3688-E40 0.059 2784-Et t 0.059 s

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                             'h                                                                                             NOTES:

l

t. CORE SPitAY SPARCER ASSOfBtX. INTEltNAL AND EXTEltNAL l "r'"TS/ PADS ARE REBIOVED FOR C!ARITY. Q i 6

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GENcclear Energ GENE-$23141 1993, Rev.1 11: 11:

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                                                                               ~       m 0.906 DETAIL 0 0.354                   e~    ~N o               /    ,

k 0.08 R , al , f

       "                             0.35 FILLST                                            30' WELD THROAT                           45{                                                                                                                                       ,

0.906 - j ~ 1.102 g-V. DrA:11 O OUTSIDE INSIDE

                                                                                   -                                                                                                                        a i                                                                                                                         a g

i: JETA11 C 5 2 Figure 2-4 '

   - _ _ _ - - - - _ _ _ _ - _ _ _ _                        __m. _ _ _   _           __

si I 3 i

                                           'Q 0.906 4y-                                                                             /'             \                            l
                                        -A
                                            \

1.102 0.08 R

                                                                                                             /                      ,   [y f  j
                                          \,            DETAIL P                                                                    'j 0.354         h N'

0.35 FILLET 0.118 ( 30' hy 45-WELD THROAT 0.060 ,

                                      \ YY     h
                                                    //,                                                JE"A::1 ?

M , 0.906 ,, OUTSIDE l INSIDE @

                                                           .                                                                                                          E E

JT?A::L J  ! l' 1

Figure 2-5 i t r

R iF i D O.060 ,

                         .                                                                                              0.188 DETAIL Q 0.354
                 ,/      ./    /                                                                                                                                l7 j    ,    j/ [                                                                                            al                                  .

f 0.08 R Ag.,J( 30' LLET HROAT 4 k, ~ N' - 0.35 FILLET 0.906 - * \ WELD THROAT ' 1.102 --Ng-OUTSIDE ._ INSIDE DEAR Q ,g n ., s J:B"A: .m :B  ! i r

                                                                                                                                                                                                  ~

Figure 2-6 t

o e m '

!                                                                                                                                                                                                            p         I iE 4

L'S 6  ! 4 OUTSIDE INSIDE i i tJ . o 30'  !

                                                                                                                                      '                                                                                t 9                                                                                !

0.094 =!= 0.500 e 9 i M e w t i DETAIL 7 )- 2 ,

                                                                                                                                                                                                              ~        ,

s Figure 2-7

R F h OUTSIDE ,/ INSIDE / , [y

                                                    /                                     0.08 R                   /                      ,
                                                    /                                                C--                                                            0.354
                                                      ./                         DETAIL R                                                         0.118
                                                    /'

[/7,h'45 s 0.060 e ' 30- Q bf

                                       \            W /l

__ \ 1.102 7_1.7a r. .,A--- M. a N. Ni . . .

                                          --           y- 0.906 e

N'/\ V  !

                                                                               -                                                                                            e E
                                                                                                                                                                            .E JrAID G                                                                                                                                  r Figure 2-8 I

El 5 OUTSIDE ./ INSIDE / , j/]n

                           /                                                                                                           0.08 R           /    ,

V 0.354

                           /'/

j' DETAIL S 0.118 m

              ' 30

[/7, i h"s 45 0.060 (\ I j

                               \ -      o2                                                                                                      DI. AI:1 S
                         --Q
                       ~

3- 0.906

                           \'f\                                                                                                                                                              n V              .

VIEW ROTATED 90'  !

                                                                                                                                                                                            .e
        .                                                                                                                                                                                   y       '

i/ /J/X DI. A:iS Figure 2-9 i e

l I 1 . GENmlear Ecergy GEVE-5231411093, Rev.1 3.0 CHEMISTRY AND FLUENCE CONSIDERATIONS 3.1 Water Chemistry History For the first decade of hot operation, Peach Bottom Unit-3 operated with relatively high primary water conductivity. As can be seen in Figure 3-1, Unit-3's arithmetic mean conductivity actually exceeded 1.0 pS/cm in 1976 and 1977. The arithmetic mean conductivity was very high and exceeded 0.4 S/cm through 1986. Subsequently, conductivity values steadily decreased, and was <0.1 S/cm (0.089 pS/cm) during 1992

and 1993. These last two year's conductivity values are considered world class performance.

The high conductivity during the first half oflife was partly due to leaking condensers and resin change out problems. However, besides the high early life steady state conductivity, there was also one relatively severe transient experienced at Unit-3 as, presented in Table 3-1, a summary of BWR fleet severe transients through about 1983. As can be seen in I Table 3-1, Unit-3 suffered at least one power resin intrusion (Incident Rank 13) during which the conductivity reached 23.6 pS/cm due to possible condensate demineralizer resin intrusion. This intrusion type ofincident results in the injection of sulfate into the RPV. j Since IGSCC initiation and propagation in sensitized austenitic stainless steel and nickel base alloys are controlled by the rate of cathodic reduction of species such as dissolved oxygen, hydrogen peroxide and/or various oxyanions, then the additional presence of a detrimental oxyanion, such as sulfate, would increase the cathodic current and thus accelerate anodic dissolution at the crack tip, i.e., IGSCC (Reference 3-1). 3.1.1 Effects ofImpurities on IGSCC An example of the effects of sulfate / conductivity on crack initiation in uncreviced material is presented in Figure 3-2. It is clear that an increase in sulfate / conductivity results in an acceleration in crack initiation as measured by the constant extension rate test (CERT) (References 3-1 through 3-4). A specific Peach Bottom example of acceleration in crack , i propagation rate (creviced) with sulfate is shown in Figure 3-3. Figure 3-3 displays June

                                                       ~

I 1986 (not included in Table 3-1) Unit-3 on-line crack monitoring data for sensitized Type 304 stainless steel. The results clearly illustrate the change in crack growth observed after i two closely linked water chemistry transients of 4-5 S/cm, i.e., increases in water 23

GENular Euray GENE SW M-10H, Rev.1 conductivity due to intrusions of demineralizer resin material (Reference 3-5). This figure demonstrates the dramatic increase in crack growth rate (2X) with conductivity. Similar on-line crack monitoring results with sulfate have also been documented in the laboratory, I l Figure 3-4 (Reference 3-6). Other anions such as chloride, carbonate, etc. have similar ( kinetic effects on IGSCC initiation and propagation (References 3-7 and 3-8). ) L l This high conductivity crack initiation and propagation acceleration factor is consistent l with the relatively high incidence ofIGSCC observed at Unit-3 in creviced Alloy 600 l I shroud head bolts (15 of 25 bolts examined cracked) and access hole covers. No cracking l of these two components has been identified in Unit-2. Both units have suffered IGSCC l of creviced safe ends. Additional documentation on the strong correlation oflGSCC I susceptibility with actual BWR plant water chemistry history for creviced BWR l components has been published (Reference 3-9). 3.1.2 IGSCC Modeling Finally, the effect of conductivity on crack propagation has also been quantified at the GE l Research and Development Center based on a "first principles" model of crack advance known as the film rupture / slip dissolution model (Reference 3-10). Predictions from the film rupture / slip dissolution model, PLEDGE (Plant Life Extension Diagnosis by GE), have been extensively compared with laboratory and 6 eld data and has provided validation of the technique. For example, PLEDGE predicts the crack growth rate in stainless steel and low alloy steel within a factor of approximately two' for a 70% statistical confidence over a range in observed crack growth rate of more than six orders of maginitude. Likewise, it provides a very reasonable mean value and can accurately bound the observed l crack growth rate in stainless steel piping and other components. Aside from piping predictions, PLEDGE has been successfully used for on-line crack growth monitoring data, safe ends (avoiding mid-cycle plant shutdowns), non-sensitized (stabilized) stainless j steels and reactor internals such as the core shroud, top guide, access hole cover and in- J core monitor housing. The PLEDGE model ofIGSCC and more recently IASCC (Reference 3-11) indicates the strong effect of conductivity on crack growth rate and by l inference crack initiation. , I

                                                                                        ~

Figure 3-5 presents a schematic estimation of Unit-3 crack growth rates as a function of conductivity using PLEDGE. Crack growth rates based on actual conductivity averages for the first ten years (0.752 pS/cm) were compared to those averages for the last two, 24

         . . .              .-                  ....r    _,-  m ,. .. . , _           ,     .     .. ,_ _ ,_. ..___ _.._                      . . _ , . .

t . . I GENuclear Emrgy GEVE-583-H1 1M, Rev.1 years. A value of 200 mV[SHE] was used for the electrochemical potential (ECP) in l these calculations. As noted in Figure 3-5, a factor of approximately nine decrease in ) l

j. crack growth rate is obtained with the unit's decrease in conductivity. Thus, crack growth j l over the past few years has been significantly reduced by proper control of water l I chemistry.

l 3.2 Fluence Considerations ) An important parameter which helps in the evaluation of the cracking mechanism is l fluence. The fluence is the time integrated flux at a particular location. Shroud peak J fluence was calculated by multiplying peak flux at the shroud location by the effective full power seconds of operation. The peak fluence in the Unit-3 shroud at the end of the next fuel cycle will be approximately 7.9x10 20 n/cm2(E>1Mev). Typically, the fluence varies with shroud azimuthallocation and elevation. Although peak shroud flux may vary significantly from cycle to cycle, available flux results are generally limited to one operating cycle per plant due to substantial resource requirements for vessel flux analysis. Shroud fluence estimates were therefore calculated based on the assumption that flux remains constant throughout the life of the plant. Further evaluation would be needed to quantify the uncertainty associated with this assumption. However, the method of determining fluence is considered to be sufficient to obtain an estimate of the overall condition of the material with respect to irradiation effects. The impact ofirradiation on core materials including crack growth rates has been studied and is discussed in References 3-11 and 3-12. l O 1 25

l

• GENE.3231411M3, Rev.1 GENanclear Eneray l

3.3 References 3-1 W.J. Shack, et al, " Environmentally Assisted Cracking in Light Water Reactors: Semiannual Report April - September 1985," NUREG/CR-4667, ANL-46-31, June 1986.

                                                                                                                                  )

l 3-2 W.J. Shack, et al, " Environmentally Assisted Cracking in Light Water Reactors: l Annual Report October 1983 - September 1984," NUREG/CR-4287, ANL-85-33, June 1985. l 1 3-3 L.G. Ljungberg, D. Cubicciotti and M. Trolle, " Effects ofImpurities on the IGSCC i l of Stainless Steel in High Temperature Water," Corrosion, Vol. 44, No. 2, February > 1988. ) i 3-4 W.E. Ruther, W. K. Soppet and T. F. Kassner, "Effect of Temperature and Ionic

,                       Impurities at Very Low Concentrations on Stress Corrosion Cracking of Type 304 Stainless Steel," paper 102 presented at Corrosion 85, Boston, MA, NACE, March                            ,

1985, published in Corrosion, Vol. 44, No.11, November 1988. l 3-5 D.A. Hale and C. G. Diehl, "Real Time Monitoring of Environmental Crack Growth in BWRs", paper 455 presented at Corrosion 88, St. Louis, MO, NACE, March l l

l. 1988.

3-6 B.M. Gordon, Corrosion and Corrosion Control in BWRs, NEDE-30637, p. 6-22, 1 December 1984. 3-7 R.B. Davis and M. E. Indig, "The Effect of Aqueous Impurities on the Stress , Corrosion Cracking of Austenitic Stainless Steel in High Temperature Water," paper l 128 presented at Corrosion 83, Anaheim, CA, NACE, April 1983.

                                                                                                        ~
                                                                                                                                  \

3-8 P.L. Andresen, "A Mechanism for the Effects ofIonic Impurities on SCC of ) Austenitic Iron and Nickel Base Alloys in High Temperature Water," paper 101 presented at Corrosion 85, Boston, MA, NACE, March 1985 3-9 K.S. Brown and G. M. Gordon, " Effects of BWR Coolant Chemistry on the

!                        Propensity for IGSCC Initiation and Growth in Creviced Reactor Internals Components," paper presented at the Third Int. Symp, of Environmental Degradation                        f of Materials in Nuclear Power Systems-Water Reactors, Traverse City, MI, August                          i 1987, published in proceedings of same, TMS-AIME, Warrendale, PA,1988.

3-10 F.P. Ford et al, " Prediction and Control of Stress Corrosion Cracking in the - Sensitized Stainless Steel / Water System," paper 352 presented at Corrosion 85, Boston, MA, NACE, March 1985. ~

                                                                                                                                  )
                                                                                                                                  \

l 1 ! 26 l

1 . GENuclear Eneray GENE 313141109), Rev.1 l

   .                                                                                                    l 3.3 References (cont'd) 3'-l1 P.L. Andresen and F. P. Ford, "Modeling ofIrradiation Effects on Stress Corrosion Cracking Growth Rates," paper 497 presented at Corrosion 89, New Orleans, LA, NACE, April 1989.

l 3-12 P.L. Andresen, F.P. Ford, and A.M. Murphy, " State of Knowledge of Radiation l Effects on Environmental Cracking in Light Water Reactor Core Materials," l Proceedings of the Fourth International Conference on Environmental Degradation l of Materials in Nuclear Power Systems - Water Reactors, Jekyl Island, GA, August 1989, NACE,1990 t i l l l l . 1 27 i

l l l GENuclear Enerv GENE-523141109), Rev.1

     ~

{ Table 3-1 Severe Water Chemistry Transients in BWRS MAX MAX 10/07/93 { COND. oH. C1 POW CATE CA?A RAnc PLANT uS/cm mm :o L!v y-m d COMMENTS REFERENCE 701%f l

       ... ............. ...... ..... ..... ... ...... .................................................. ............. ....                          l 1 AG              95.0     45      1CO P 780307 CONO 0EMIN RE5[N BLEE0T4 ROUGH                               PC&RT 821.0A01       35        l 2 AG              88 0                          P 8C0802 CONDENSATE DEMIN RE5th INTRUSION                    EPRI NP 4134         46        I I

3 Al, 84.0 3.2 14500 P 120901 CONDENSER LEAK DEMIN 00EPLETED PC&RT 82 LOA 01 6 4 OZ 72.0 560 P 660820 PC&RT 82 LOA 01 1 ! 5 AG 10.0 4.6 198 P 711116 CRUD & CONDENSATE DENIN RE5IN INTRU510N PCLRT 82 LOA 31 30 , 6N 54.0 3.8 P 740804 RE51N BEAD INTRU$10N

• NEDE 13405 13 l f 7 AB 40.5 3.9 P 740608 AIR / AIR RE51N N!1TURE INJECTED INTO R FROM RVCU PCLRT 82LDA01 11 l 8N 33.0 4.0 P 740426 RESIN SEAD INTRUSION
• NEDE 13405 to 9 AC 30.0 P 710903 HIGH CONOUCTIV!TY WATER IN CST PCLRT 82 LOA 01 3 i l 10 OZ 28.5 P 661130 PC&RT 82 LOA 01 2

                                                                                                                                                      )

11 8 25.6 4.1 50 P 170601 RESIN INTRU510N PCM T 82LCA01 28 12 8 25.0 2500 o 810412 CONDENSER LEAK PMET 81 688 45 52 13 PEACH BOTTOM 3 23.6 P 800205 P055!8LE CONDENSATE DENIN RE51N INTRU510N EPRI NP 4134 80 i 14 OZ 23.0 3000 P 790407. LEAKAGE OF C00 LING WATER INTO RPV VIA CORE SPRAY PC&RT 82 LOA 01 42 l l 15 W 23.0 30 P 730406 AIR INJECTED INTO Rx FROM RWCU PC&RT 82LCA01 7 16 AG 22.0 P 771212 CONO DENIN RES!N INTRUSION PC&RT 82 LOA 01 32 17 K 21.0 4.6 2500 P 820428 TRICH 1.0R0 ETHANE FRON RADWASTE AND CST PC&RT 82 LOA 01 53 18 AG 20.0 4.5 o 821004 POSSIBLE CONDENSATE DENIN.RE5!N INTRU$10N EPR! NP 4134 94 19 F 17.0 P 801001 CONDENSER TU8E LEAKS PC&RT 82 LOA 01 49 20 C 14.0 P 750605 RWCU OUT OF SERVICE PC&RT 82 LOA 01 20 l 21 T 13.8 4.7 100 P 781110 ORGANIC INTRUSION VIA CONDENSATE, DECON DETER /0!LS PC&RT 82LDA01 39 22 A8 13.5 P 740925 HIGH COMO WATER PC&RT 82 LOA 01 14 23 AB 13.0 100 P 803428 UNKNOWN (LONG$ HUT 00WN) EPRI NP 4134 81 24 T 12.1 P 78022: !!WCU RE51N INTRU$10N PC&RT 82 LOA 01 34 25 0 12.0 P 750702 CON 0!NSER TUBE LEAK PC&RT 82LDA01 21 26 0 12.0 4.8 50 P 761025 RWCU RES!N TRAP, RWCU IN0PERA8LE- PCMT 82 LOA 01 25 11.8 P 800412 ORGANIC INTRu$10N EPRI NP 4134 82 27 T 28 0 11.5 4.8 60 P 750127 RWCU RESIN INTRU$10N ,, PC&RT 82LDA01 17 29 AB 11.3 4.7 p 830106 P0$518tE CONDENSATE DENIN RE51N INTRU51M EPRI NP 4134 95 10.8 4.5 50 P 750601 RE51N FRON FLUFFING CONDENSATE OF/0 PCSRT 82LDA01 19 30 8 PC&RT 82LDA01 8 31 AG 10.6 4.5 100 P 730507 RWCU RESIN INTRUSIM 10.0 P 741208 CON 0(NSER LEAK PC&RT 82LDA01 15 32 0 10.0 p 820418 RWCU RESIN INTRUSION EPRI NP 4134 92 33 AG EPRI NP 4134 74 34 H 9.2 7.4 57 P 780211 CONDENSATE DENIN RE51N INTRus!0N EPRI NP 4134 63 35 8 8.2 4.3 50 P 151210 WASHOUT OF INPURITIES FROM T1st81NE 8.0 5.0 500 P 790516 PMET 81 688 45 44 36 8 p 810411 RWCU RE$1N INTRU$10N EPRI NP 4134 86 37 8 7.5 89 38 C 7.1 100 P 811010 DECOMP 0$1 TIM 0F RA0 WASTE REllNS DUE TO HOT WATER [ PSI NP 4134 EPut NP 4134 84 39 C 6.5 P 810210 CAUSTIC INTRU$10N VIA CONDENSATE STORAGE EPRI NP 4134 57 40 K 6.2 4.8 20 P 741118 $USPECTED REllN INTRU$10N EPal NP A134 54 41 0 5.8 4.5 50 P 730812 $USPECTED REstu INTRU$10N 5.6 P 700123 RESIN INTRUSION WNEN C/D RET 18tNED TO SERVICE JNs QC $30717 98 42 AA EPRI NP 4134 64 43 8 5.4 4.7 50 P 151218PR08A8L'tRWCURE$1NINTRU$10N EPRI NP 4134 85 44 8 5.1 p 810220 CRGANIC INTRUSION VIA RA0 WASTE 68 P 770727 CONDENSATE DENIN RES!N INTRU5 ION, ANION RICH EPRI NP 4134 70 45 H 5.1 EPRI NP 4134 67 46 0 5.1 4.8 50 P 760806 CONDENSATE DEMIN REllu INTRU$10N 28

• GENuclear Enugy GENE-523141 1093, Rev. !

l l Table 3-1 Severe Water Chemistry Transients in BWRS l l MAX MM 10/07/93 CONO. DM. C1 POV OATE DATA RANK PLANT uS/cm. mm D00 LEV y-m d COMM(NTS REFERENCE POINT 47 C 5.0 100 P 750309 POS$!BLE CONCENSATE DEMIN REstN INTRUSION EPRI NP 4134 79 48 Q 4.9 49 50 P 160522 $USPECTED RE$lN INTRUSION EPRI NP 4134 66 49 T 4.5 5.0 50 P 781227 ORGANIC INTRU$!CN VIA CONDENSATE SYSTEN EPRI NP 4134 17

        $0 g                   4.3     4.9       48 P 760221 $USPECTED RESIN INTRU$10N                             EPRI NP 4134         65 51 K                   4.1     5.1       80 P 141015 RWCU RESIN INTRU$10N                                  EPR[ NP 4134         56 52 AL                 3.3     5.4       50 P 170126 !MPROPER, RINSE OF CONOENSATE DEMIN                   EpR[ Np 4134         68 53 C                  3.3     5.2       38 P 750309 POS$18LE RE$lN INTRU110N                              EPRI NP 4134         59 54 8                  3.2                    o 810715 RWCU RES!N INTRU$10N                                EPR! NP 4134         88 55 $                  3.2     4.7     495 P 780131 RES!N INTRUSION                                        EPRI NP 4134         73 56 H                  3.0     5.6       96 P 750902 $U$PECTED RESIN INTRU$10N                             EPR! NP 4134         61 57 8                  2.9     5.4       50 P 751126 PR08ABLE RWCU RES!N INTRUS!ON                         EPRI NP 4134         62 58 T                   2.8    5.2       65 P 770912 !MPROPER RIN$E OF CONDEN$ ATE DEMIN                   EPR! NP 4134          71 59 8                   2.7    7.6            P 750626 RE$!N INTRU$10N                                     EPRI NP 4134          60 60 T                   2.3                   P 800824 ORGANIC INTRU$10N                                   EPRI NP 4134          83 61 T                   2.2    5.5       50 P 790108 SV$PECTED ORGANIC $ IN CONDEN$ ATE STORAGE            EPRI NP 4134          78 62 T                   1.8    5.4      355 P 781208 CONDENSATE DEMIN RE11N INTRus!0N                       EPRI NP 4134         16 63 AC                  1.4    5.6       83 P 741125 VALVING ERROR DURING RESIN TRANSFER                    EPRI NP 4134         58      1 64 H                   1.4    s.1       38 P 780112 CONDEN$ ATE DEMIN RESIN INTRU$10N                      EPRI NP 4134         12 65 AJ                  1.4                    P 750906 $USPECTED FLOC / FILTER A!D/$URFACT FROM RAD WA$fE JMS QC 930717 99 66 0                   1.1     5.6      30 P 770225 $USPECTED RES!N INTRU$!ON                              EPRI NP 4134         69 67 H                   1.1     8.8       72 P 780511 CONDENSATE DEMIN RE$1N INTRUSION                      EPRI NP 4134         75      ,

68 W 1.0 P 811030 GTLCOL INTRU$10N VIA RADWA$TE EPR! NP 4134 90 f 69 PEACH 80TTOM 2 1.0 P 810622 O!L INTRU$10N INTO HOTVELL EPRI NP 4134 87 725 P 711113 HIGH FEE 0 WATER CONOUCTIVITY PCMT 82 LOA 01 4 70 A$ 71 8 540 P 780129 PCMT 82LDA01 53 72 8 1200 P 160708 CONDENSATE ST3 TEM MOMENTARILY BTPA$$E0 PCMT 82LDA01 24 600 P 750103 RWCU OUT OF $ERVICE PCMT 82LDA01 16 13 AR 74 8 641.0 3.5 87000 $ 790426 COOLING WATER INGRE$$ FROM RML, RPV H2O TO HOTVELL PCMT 82LDA01 43 423.0 3.2 $ 740801 t.CIO INTO RPV FROM DEMIN STORAGE TANK PCMT 82LDA01 12 75 8 76 H 140.0 $ 760519 PCMT 82LDA01 22 45.9 3.8 244 $ 761103 TORUS WATER PIMPED INTO RPV PRIOR 70 STARTUP PCMT 82 LOA 01 26 77 7 13.3 1800 $ 160520 PCMT 82LDA01 23 78 8 13.0 $ 780801 LEAK IN RMt HEAT EXCHANbER PCMT 82 LOA 01 38 19 A 12.9 $ 770917 RWCU OUT OF $ERVICE PCMT 82LDA01 29 80 8 12.1 $ 800815 PCMT 82LDA01 47

 .        N1 T
                                                        $ 730403 RWCU OUT OF SERVICE                                  PCMT 82 LOA 01        9 82 AL                11.6 11.2                    $ 800822                                                      PCMT 82 LOA 01 de 83 8 PCMT 82LDA01         51 84 0                  11.2                    $ 801219 EPRI W 4134         91 85 F                  10.5                    $ 820427 POS$!8LE CRGANIC IRTRU$!0N EPRI W 4134         31 86 8                 10.5     5.6      140 3 780923 RWCU RE$1N*    INTRU$10N PCMT 82 LOA 01      18 87 H                 10.3                    $ 750405 RWCU 0UT OF $ERVICE PC8RT 82 LOA 01       5 88 AS                10.0              730 $ 720604 DEPLETED RWCU DEMIN EPRI W 4134         55 89 0             . 5.0      5.5      60 $ 740829 CON 0ENSATE DEN!N RES!N INTRUSION EPRI W 4134         97 90 AX                 4.5      5.2     220 $ 830505 ORGANIC INTRUSION VIA RADWA$TE 4.2     5.3     600 $ 781110 ORGANIC INTRUSION VIA CONDENSATE, DECON DETER /0!L$ PCM.T 82 LOA 01 34 91 8                                                                                                                            93
                                                          $ 820900 GYLCOL INTRU$10N VIA RADWA$TE                       EPRI NP 4134 92 AP                  1.0 29

1 . GENuclear Energy GENE.523.H11093, Rev. I l l Table 31 Severe Water Chemistry Transients in BWRS ex xx COND. pH. C l. POV OATE 07/93 RANK PLANT uS/cm min poo (Ey y.m.d ' COMM

 ... ............. ......     ... ..... ... ,,,,,, ,,,,,,,,,,,,,,,,,,,,,,,,,ENTS ,,,,,,,,,,,,,,,,,,,

REFERENCE POINT 93 F 700 $ 800305 94 T PCMT 82LCA01 45 1300 1 790329 CONDENSER LEAK CONDEN$ ATE SYPASSED. PCMT RVCU QUT 82LDA01 41 95 0 500 $ 801017 PC M T 82 LOA 01 50 96 AC 683 5 770309 PCMT 82 LOA 01 27 97 ( 5 830213 GTLCOL INTO RADWA$TE OETECTED PRIOR TO COND STOR 96 EPRI NP 96 T 1200 5 790316 CONDENSER LEAK, CONDENSATE DEPLETED. Cl INTO CST PCMT 82LCA01 40 99 8 800 $ 771206 RVCU OUT OF SERVICE PCMT 82 LOA 01 31 NOTE: 8WR$ RANKE0 IN THE FOLLOWING ORDER: 1

1. POWER (P) 04 $HUTDOWN ($) l

2. CONDUCT!VITT l OTHER NOTES: * . RE$[N 8EA05 PROVICE LONG TERM LOW pH l l

O b 30

                                                               ~,-r -

Arithmetic Mean Conductivity a Peach Bottom 3 f& Conductivity, pS/cm , 1.4 - 1.2 - l 1

                                                    ' Peach Bottom 3 all years = 0.536 3

0.8 - i O.6 - 1 0.4 - o h O.2 - _. 4 $

                                                 '        '      '         '      '                E 0                                                                 90      92         94     3 74      76   78       80      82       84       86     88 Fuel year Figure 3-1 Peach Bottom 3 Arithmetic Mean Concluctivity

Effect of Concentration and Conductivity on IGSCC Initiation - FS Type 304 l t p Acceleration Factor '

                                                                                                                                                                                                                                                                                  $       i 5

O ANL Data V EPRI Data 4 - t i O M 3 - V 2 - O , conductivity (usicm) e,

                                                                                                                                                                                            -                                                        -   E                        h,
                                                                                                                                                 ,1'
                                                                                                                                                             !',       'i" f ,* P,'i , , , ,,,,                                                     i, , , , , , , ,

10 , , , , , , , , , 10000 $ 10 0 1000 I 10 Sulfate (ppb) {9 l Crack initiation data based on CERT FS Type 304 Figure 3-2 Effect of Concentration and Conductivity on IGSCC Initiation l

Peach Bottom 3 Response to June 1986 Water Chemistry Transient l 7 D Crack length, mm g 19.16 Resin Resin Intrusion 1 Intrusion 2 ~ 19.14 - 19.12 - -

                                                                                                           ,.. , ' , .       da/dt - 1.09 mm/y y                                                                   _.

19.1 - - da/dt = 0.55 mm/y 19.08 - o I i 'l I I I 19.06 i 210 0 2200 2300 2400 2500 2600 2700 2800 g e t Hours ) s Figure 3-3 Peach Bottom 3 Response to June 1986 Water Chemistry Transient I e

1 MM IN,. SCL 1 CONSTANT LO AD - - - - - -- - ~ ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . , 20.8 - 0.82i_  !

1.0 pS/CM

Na2SO4 k 0.45 pS/cm Na2SO4

                                                                                                                                                                                                                                   <0.1 pS/cm                                                                                                                                             .;,,
                                   ,". O 5 pS/cm                            <0.1 pS/cm                                                                                                         z                                                                                                                                                                                          R 20.5             -       0.81     --

9 e  ; g t  %  !

> no z L

                                      -                              z                                                                                                                         <                                                                                                                                                                                           s u>                                                                                                                         1                                                                                                                                                                                          d

_ _s t-o 20.3 - 0.90 -- _ --- 20 PPD OXYGEN --- p ----200 PP8 OXYGEN --- --

                                                                --      -200 PPS OXYGEN -
                                                                                                                                    -------~~----<                                                                                                                                                                                                                                                    ,

s 20.2 - 0.73 -- I i 19.5 - 0.78 - i

                                         ~                                                                                                                                                                                                                                                                                                                                                             ,
                                           -                                                                                                                                                                                                                                                                                                                                                           i e

u -  ; 4 - , i 13.6 - 0.77 4-  ! I I ann

_ - - - - t 13.3 - 0.76 --

p - 1.5 pM/h l 1.1 pM/h O.1-0.03* pM/h O 09 pM/h (4.1-1.2* pinJh) (80* pinJh) , l 13.s pin.th) (46 pia./h) 4 t'1 l 4 0.75 c. 19.0 -

                                                                                                                                                                              .                                                                                                                                                                                                             E t

i

                                                                                                                                                                                                                                                                                                                                                             '                              E I

2000 2500 ,$ 18.8

• O.74 -

1000 1600  % O 500 g HOURS 'NO COMPLEANCE UNLOADING Crack Length Versus Time, Sensitized Type-304 Stainless Steel 286*C Oxygenated

 ~

Figure 3-4 ~ Water Environment, K = 28.6-30.5 MPadm (26-28 ksi 4Td) t

e i a GENE PLEDGE Model Prediction for PB-3 Sensitized Type 304 Crack Growth Rate l@ . Crack Growth Rate, in/h 1975-1984 200 mV 100 mV c ni O mV

5. -

                                                                                     -100 mV M

1992-1993 ' 1.000E-05 -

                                                                                     -200 mV
                         -                                                           -300 mV t

1.000E-06 - h v y i

                           -                                                                       ~

1.000E-07 0.6 0.7 0.8 0.9 s 0 0.1 0.2 0.3 0.4 0.5  := . Conductivity, pS/cm [ PLEDGE: 15 C/cm2, 20ksi/in Figure 3-5 GENE Pledge Model Prediction for Peach llottom 3 Type 304 Crack Growth

       .-        - . _    -          - _ . - -      . -        - . - .       - - - .          . . - - . . - . . ~ . _ . - - - -

l . . l

• GEN:elear Exergy GENE-5231411093, Rev.1 4.0 IN-VESSEL VISUAL INSPECTION This section summarizes the IVVI results of the Peach Bottom Unit-3 core shroud. IVVI of welds H1 through H8 were performed during this outage. The IVVI included both inside surface and outside surface examination. Figures 1-2 shows the indications associated with the H3 and H4 welds. Only a few short indications were observed on the .

outside surface of H1 and H4. Circumferential indications were observed on the inside surface associated with vertical weld V3 (See Figure 1-2). It should also be noted that the , area adjacent to the H9 weld was visually inspected as part of access hole cover (AHC)  ; l inspection at this cutage. The inspection did not reveal any indications. The H9 weld in the vicinity of the access hole cover is considered a higher stressed location and therefore l l these IVVI results are considered to provide a reasonable assessment of the overall t condition of the entire H9 weld. l All indications associated with the H3 weld inside surface were in the HAZ of the shroud l cylinder. No indications were found in the ring. The indications near H3 were all circumferentially oriented. As can be seen in Figure 1-2, most of the indication length is

located between the azimuth of 146' and 360'.

The indications observed at the H4 inside surface HAZ were a mixture of circumferertial and axial indications as shown in Figure 1-2. Circumferential indications were observed emanating from the V3 weld on the inside surface. Eight indications were observed grouped together with a spacing of ! approximately 2" between indications. All other IVVI venical scans found no indications. I On the outside surface a limited number of short indications were observed associated with H1 and H4. Table 4-1 is the IVVI plan which indicates the original planned inspections. Due to the l observation ofindications at H3 and H4, the inspection scope was expanded and is also described in Table 4-1. A summary of the IVVI results is shown in Table 4-2. See Figure ' l-2,1-3 and 1-4 for further details. I 36 l

GENutear EmaY GENE 53314110H, Rev.1 i . Table 4-1 j; Peach Bottom Unit 3 Core Shroud Exam Plan (3R09) ! ORIGINAL PLAN (Prior to issuance of SIL 572) i

1. Perform simple examination "ID" at (8) cell locations of the "H3" and I "H4" welds.

3 1

2. Perform sample examination "OD" at (8) locations in the high flux areas  ;

j at welds "H1, "H2", and "H5" l I ~ EXPANDED PLAN (Following identification ofindications on the "H3" and

"H4")

< I j 1. Perform 100% examination of the "H3" and "H4" welds from the ID. l ]

2. Perform 100% examination of accessible areas of the "H4" weld from the j OD. ,

1 j 3. Perform examinations of the "H3" weld, "OD", where cracks were not

identified from the "ID" 1

j 4. Perform an examination of the "H3" weld "OD", including significant

corresponding areas of cracking identified on "ID".

[ 5. Perform a sample examination on the "OD", at (8) locations,of the "H6" ! weld. t i j 6. Perform a sample examination of the "OD" at (2) locations of the "H7" i and "H8" welds

)           7.          Perform an examination of(l) vertical weld between the "H3" and "H4"
!                       weld.

8. Perform a sample examination of the plate to include:

(1) 8" area at the vertical weld. (1) 8" area between "H3" and "H4" welds. ' (1) 2" area between "H3" and "H4" welds.

  =

NOTE: Consideration was given to high neutron flux, stress, and repair areas for selection of the sample locations 37

GENuclear Energy GEhE-333-141 1993, Rev.1 Table 4-2 Summary orIVVIIndications Weld Inside Surface Indications Outside Surface Indications HI N/A 1 short vertical H2 N/A None H3 Circumferentialin Shroud None Cylinder HAZ H4 Circumferential and Axial 2 shon vertical HS N/A None H6 N/A None H7 N/A None H8 N/A None V3 8 short circumferential N/A PLATE None N/A Ob e 9 e g e** 38 l j

_, - _. ~ .. j , GEN: clear Eneray GENE.5231411M3, Rev. I 5.0 FLAW EVALUATION l This section provides the flaw evaluation and application of the screening criteria to the Peach Bottom Unit-3 indications. Included in this section is the structural analysis, allowable flaw size determination, and screening criteria. l 5.1 Structural Analysis l This section describes the details and the results of the structural analysis performed to , determine the allowable flaw lengths. The structural analysis consists of two steps: the l determination of axial and circumferential stress magnitudes in the shroud, and the l i calculation of the al!owable flaw lengths. Both the fracture mechanics (LEFM) and limit I l load methods are used in the calcul.ation of allowable flaw lengths. ' 5.1.1 Applied Loads and Calculated Stresses The applied loads on the shroud consist ofinternal differential pressure, weight and ) seismic. The seismic loads consist of a horizontal shear force at the top of the shroud and i an overtuming bending moment. The shear force produces a shear stress ofinsignificant magnitude, and is not considered. The bending moment stress at a shroud cross-section varies as a function ofits vertical distance from the top of the shroud. Because of the inherent ductility of the material, residual stresses and other secondary stresses do not affect structural margin. Thus, they need not be considered in the analysis. The magnitudes of the applied loads were obtained from the seismic stress analysis and . system information reports. The nominal shroud radius and thickness (2.0 in.) were used to calculate the stresses from the applied loads. The stresses are essentially based on the strength of materials formulas. Since the bending stress due to seismic shear force varies with the elevation of a location, two conservative values of this stress were calcolated: one applicable to shroud sections above the core plate (H1, H2, H3, H4, and HS) and the other for sections below the core plate (H6, H7 and H8). Figure 5-1 shows the weld designation and relative locations in the shroud. 1 l l l 39 s

GENuclear Ewgy GENE 3231411093, Rev.1 l Table 5-1 shows the calculated seismic stress magnitudes for both the upset (Design Earthquake - DE) and faulted conditions (Maximum Credible Earthquake - MCE). The appropriate pressure differences for the upset and faulted conditions are shown in Table 5-2. l Table 5-1 Seismic Axial Stresses at Shroud Welds i I Weld MCE Stress (ksi) Moment 1 Designation (ft-kips) MCE DE HI 1104.7 0.18 0.08-H2 1438.6 0.23 0. I 1 H3 1479.1 0.27 0.13 l H4 2995.8 0.54 0.24 ! H5 4583.8 0.83 0.37 i ! H6 4679.7 0.90 0.40  ! H7 5697.6 1.10 0.49 l l H8 6749.7 1.30 0.58 I l l l- Table 5-2 Pressure Differences  ! l Pressure Differences (psi) Component Faulted Condition Upset Condition  ! l Shroud Head and 32.9 14.12 l Upper Shroud l Core Plate Support Ring 54.8 35.68 , l and Lower Shroud The structural analysis for the indications uses two methods; linear clastic fracture mechanics (LEFM) and limit load analysis. Both the limit load and the LEFM methods were used in determining the allowable flaw sizes in the shroud. Since the limit load is  ; concerned with the gross failure of the section, the allowable flaw length based on this approach may be used for comparison with the sum of the lengths of all the flaws at a cross-section. On the other hand, the LEFM approach considers the flaw tip fracture toughness and thus, the allowable flaw length based on this approach may be used for , i

             . comparison with the largest effecti ve flaw length at a cross-section. The technical approach for the two methods is describedblow.                                                                                                      ,

l 40 l t . . _ . _,, ._-.-.. _. - . _ _ _ - . _ _ _ - _ _ . - __ _ - . . - _

l..-. . GENuclear Eneray GENE-523-1411093, Rev.1 5.1.2 Fracture Mechanics Analysis The shroud material (austenitic stainless steel) is inherently ductile and it can be argued that the stmetural integrity analysis can be performed entirely on the basis oflimit load. In fact, J-R curve measurements (Figure 5-2) made on a core shroud sample taken from an overseas plant having higher fluence (8x1020 n/cm2) showed stable crack extension and l ductile failure. The ASME Code recognizes this fact in using only limit load techniques in i Section XI, Subsubarticle IWB-3640 analysis. Nevertheless, a conservative fracture mechanics evaluation was performed using an equivalent Kjc corresponding to the material JI c. The Kjc for the overseas plant shroud was approximately 150 ksifm. Use of l this equivalence is conservative since: j i) The calculated fluence for Peach Bottom Unit-3 is lower than that for the overseas l plant from which J-R curves were obtained. i L ii) The J-R curves show Jmax values well above the JI c, confirming that there is load 1 l capability well beyond crack initiation (See Figure 5-2).

                                                                                                                                 )

l Using the ASME Code safety factor of 3, which is applicable for normal and upset conditions of pressure boundary components, the allowable K cI value becomes 50 ksiVin. I l For faulted conditions the allowable K Ic is 107 ksiVin using the ASME Code safety factor l of 42. For the analysis presented here, the LEFM analysis is confined to the H4 weld and l above. The fluence corresponding to welds at and below the core plate elevation is an l order of magnitude lower and the associated fracture toughness is comparable to that of i the unirradiated material. For those locations, limit load analysis is used. . , I An additional consideration that applies only to the fracture mechanics analysis is the  ; question, "When is a flaw independent of an adjacent flaw?". The ASME Code proximity rule considers how flaws can link up and become a single flaw as a result of proximity. However, even when two flaws are separated by a ligament that exceeds the criterion, they may not be considered totally independent of each other. That is, the flaw tip stress intensity factor may be affected by the presence of the adjacent flaw. This can be

accounted for by using the finite width correction factor for a flaw in a finite plate. For a l through-wall flaw in an " infinite" plate, the stress intensity factor is

K = oV(xa) 41

_. _ _ . . _ _ . _ ..m _._ . . _ . _ . - _ _ _ _ __ _ . _ _ _ . . . . _ _ . _ _ _ > . l GENucleu Entray GENE 323141-1093, Rev.1 For a finite plate, the K value is higher as determined by the finite width correction factor, F. In this screening evaluation it is assumed that the plate is " infinite" if the correction  ; factor F is less than 1.1. As seen in Figure 5-3, if the width of the plate exceeds 2.5L (or a/b less than 0.4), then there would be no interaction due to plate end edge effects. If this l same condition is applied to two neighboring flaws, then there will be no interaction I between the two indications if the tips are at least 0.75(Ll+L2) apart. If the distance l between indications is greater than 0.75(Ll+L2), then they are considered as two separate l flaws. However, if they are closer, for the purpose of fracture analysis, the equivalent flaw length is the sum of the two individual flaws. l ( 5.1.3 Limit Load Analysis i A through-wall circumferential flaw was assumed in this calculation. Limit load calculations were conducted using the approach outlined in Subsubarticle IWB-3640 and l Appendix C of Section XI of the ASME Code. The flow stress was taken as 3Sm. The Sm value for the shroud material (Type 304 stainless steel) is 16.9 ksi.at the normal operating temperature of 550*F. 1 i l Safety factors similar to that used in the ASME Code (2.8 for normal and upset and 1.4 for emergency and faulted) were used in the analysis. The highest seismic stress was used for the limit load calculations and is shown in Table 5-1. Similarly, the highest axial l pressure stress corresponding to the lower shroud was used. Thus, the analytical results l are applicable for all welds since limiting values are used. l I O 42 l

 . . _ .      _   _ _ . _ _ _         _..____._.__.__.______-...____..___.._._m...                                               . . . _ . _ _ _ . _ . _ _ ,

t GENuclearEnergy GENE.523141-1H3, Rev,1 ! 5.2 Allowable Through-Wall Flaws Allowable throunh-wall flaw sizes were determined using both fracture mechanics and limit load techniques for both circumferential and axial flaws. It should be emphasized 2 that the allowable through-wall flaws are based on many conservative assumptions and are l intended for use only in the screening criteria. More detailed analysis can be performed to justify larger flaws (both through-wall or part through when measured flaw depths are j available). However, cince the intent of the screening criteria is to determine when additional evaluation or NDE characterization is needed, a conservative bounding

                                                                                                                                                             )

approach is utilized. l l 5.2.1 Allowable Through-Wall Circumferential Flaw Size Both the LEFM and limit load methods were used to evaluate the allowable through-wall flaws. Above the core plate, LEFM and limit lead analysis methods were used. Since this is a screening criteria, single allowable flaw size criteria (limiting location) was used for all weld locations. It should be noted that the H7 and H8 welds involve Alloy 600 which has higher Sm values and therefore has higher limit load capability. Fracture Mechanics Analysis The total axial pressure and seismic stress corresponding to the upset condition is 0.61 ksi, l and 1.39 ksi for the faulted condition. Using the ASME Code safety factors for fracture analysis, the faulted condition is limiting. l To determine the allowable flaw size based on LEFM rnethods, the conserva ively estimated irradiated material fracture toughness K cI value of150 ksiVin was used. Applying a safety factor of 1.4 for the faulted condition, the allowable KI of 107 ksiVin

                                                                                                                                                             ]

was obtained. The allowable flaw size was calculated using the following equation: i KI= Gm *o*V(na) where Gmis a curvature correction factor as defined in Figure 5-4 (Reference 5-1), o is the axial stress, and 'a' is the half flaw length. The allowable through-wall circumferential - f'aw length (2a) was determined as = 344 inches. 43

l . GENukeFmay GENE-H3-M.M3, Ren 1 l Limit Load Analysis l A through-wall circumferential flaw was assumed in this calculation. The limit load calculations were conducted using the approach outlined in Subsubarticle IWB-3640 and Appendix C of Section XI of the ASME Code. The flow stress was taken as 3S m

                                                                                                                    . The Smvalue for the shroud material is 16.9 ksi at the normal operating temperature of l

550*F. The suesses for the limit load analysis for the upset condition consisted of an axial force stress of 0.71 ksi, and a bending moment stress of 0.49 ksi. Similarly for the faulted condition, the axial force stress was 1.21 ksi, and the bending moment stress was 1.1 ksi. The allowable flaw length was approximately 430 in. including l the ASME Code, Section XI safety factors. 5.2.2 Allowable Through-Wall Axiti Flaw Size l Fracture Mechanics Analysis The allowable axial flaw size is governed entirely by the pressure hoop stress. Similar to the circumferential flaw case, the allowable axial flaw size was determined assuming a l through-wall flaw. For a through-wall flaw oflength 2a in the shroud, the applied stress intensity factor is given by:

                                                                                                        ~

K = M

• oh
• 4(xa) where M is the curvature correction factor. M is given by: ,

M = Gm + Gb (Figure 5-5, from Reference 5-1) In the above expression, the allowable flaw length,2a, can be determined by equating the calculated K to the fracture toughness divided by the safety factor of 3. The hoop stress l i.s 1.85 Ivi and the allowable K = 150/3 (where 150 ksiVin represents a conservative estimate of the material toughness and 3 is the safety factor). The allowable flaw length was conservatively determined to be 2a = 59 in. l i 44

_ _ . __ _ . . _ . . _ _ ___ _ _- - - _ _ . - _ - - ~ _ . - . _ - _- . _ . - - - _ . - - - i

• GENesclear Eneray GENE-313141-1M3, Rev.1 Limit Load An alternate approach to determining the allowable flaw size is to use limit load .

techniques. The allowable flaw length is given by the equation: f 1 Ch " Of/ (M1

• SF) l l where Mi is a curvature correction factor (which is a function of the flaw length l

(Reference 5-2)), or= 3S m is the flow stress, SF is the safety factor of 2.8 for upset l conditions, and ch = the hoop stress corresponding to the upset AP of 35.68 psi. The allowable flaw length based on the limit analysis is 200 in, which exceeds that determined by LEFM. Thus, the allowable axial through-wall flaw length is 59 in. l l l 1 l I S a e

• 45 e

.- . 1 GENuclear Energy GENE-523.H1-1M3, Rev.1 S.3 Screening Criteria The determination of the allowable through-wall flaws has been described in Section 5.2. l The objective was to use the at:owable flaw size as the basis for the screening criteria. Since the screening rules represent the first step in the evaluation, they are by definition conservative. If the criteria are exceeded, the option of doing further detailed evaluation or performing additional NDE remains. The effective flaw lengths (Lleg L2 e g etc.) determined by combining indications using the proximity and interaction rules, are used in the comparison with the allowable flaw sizes. The detennination of effective flaw sizes are discussed in detail in Appendix A. The allowable through-wall flaws were:

• CircumferentialFlaws

                        - 344 in. using LEFM
                        - 430 in. using limit load e AxialFlaws
                        - 59 in. using LEFM
                        - 200 in. using limit load A conservative approach in developing the screening mle is to include both the LEFM and limit load analysis. For axial flaws, the allowable flaw length based on the LEFM controls, and the screening limit is 59 in.

For circumferential flaws the fracture mechanics based limit for a single flaw is 344 in. This in itselfis not sufficient since there could be several flaws (each less than 344 in.) in a circumferential plane that cumulatively add up to greater than 430 in. (the allowable circumferential flaw size based on limit load analysis). Thus, the cumulative flaw length should be less than 430 inches. While this fully assures the ASME Code margins, an l additional conservatism is included in the screening. This states that the cumulative l flaw length cannot be more than 430/4 = 107.5 in. In any 90 degree sector of the shroud. This is a conservative restriction that assures that long continuous flaws are not l admissible. With the provision that the cumulative flaw length cannot exceed 107.5 in. in l any 90* sector of the shroud, this criterion becomes more limiting than the fracture i mechanics limit of 344 in. The approach used here for the 107.5 inch limit for circumferential flaws is to assume a template with a moving window equal to the 90* l sector.' The cumulative length of flaws that appear in the window should be less than 46 i 1

• GENuclear Energy GENE 313-1411H3, Rev.1 107.5 in. A similar restriction based on limit loads is not needed for axial flaws since they are associated only with circumferential welds and are unlikely to be aligned in the same ne.

Ine removal or reduction of the factor of 4 isjustified if the IVVI results indicate that the remaining ligament is spread around the circumference of the shroud circumference. Evaluation of the IVVI results indicate that this is the case for the Peach Bottom Unit 3 shroud (See Figure 1-2 and 1-3). It should be noted that when considering LEFM based evaluations, the crack interaction criteria described in Appendix A, must be applied in comparing against the allowable lengths. For example, the adjacent flaws where the spacing S is less than 0.75 (L1efr+ L2 eg), the length L=L1efr+ L2 eft si used for comparison with the LEFM based allowable flaw length. e e

• 47 1

1 GENuclear Energy GENE 523-141-1M3, Rev.1 l 5.4 Summary of Screening Criteria l t l The screening criteria is schematically shown in Figure 5-6. The first step is to map the flaw indications observed by IVVI. Next the proximity rules are applied to the flaw map to develop effective flaw lengths (Appendix A provides the details for determining effective lengths). The results of the effective flaw lengths are also mapped. For axial flaws located in a vertical plane, two neighboring flaws must be summed if S < 0.75(L1efy+L2 eg). If the longest resulting flawis less than 59 inches, then the screening limit is met for axial flaws. For circumferential flaws, all flaws are summed in any 90* sector using a template. The total flaw length in the 90 window must be less then 107.5 inches to meet the screening criteria. The next sitep is the LEFM. based comparison using the interaction criteria. If S <0.75 (L1efr+L2eg), then the length L = L1efr+ L2egshould be compared with the LEFM limit of 344 in for circumferential flaws. If significant ligament remains around the shroud circumference, the factor of 4 may be removed or reduced. The application of this factor would be considered conservative since the presence of the ligament around the circumference assures that extremely long indications are not present. The removal or reduction of the factor of 4 would be l considered as part of the "Further Evaluation" box in Figure 5-6. g # i

                                                                   .                                                                         I I

48

GENulee Ewy GENE-m.m.1m, Rev.1 5.5 Application of Screening Criteria The screening criteria was used to evaluate the indications found by IVVI. The structural integrity of the core shroud is assured if the screening criteria are met for all of the indications. All axial effective indication lengths are significantly less than the allowable flaw size based

either on LEFM or limit load methods. Thus the axial indications seen by IVVI are acceptable per the screening criteria developed for Unit-3.

The effective indication lengths were determined for all of the circumferential indications as shown in Figure 1-2. The calculation took into consideration the detailed geometric information such as each indications length, the azimuth of each crack tip and the spacing between indication planes in order to properly determine the effective length. The resulting effective indication lengths were then compared against the allowable indication length. Since all effective indication lengths satisfy the screening criteria, the structural integrity of the Unit-3 shroud is assured for the next two year cycle with power rerate conditions. As mentioned in Section 5.3, for the indications observed in the Peach Bottom Unit 3 shroud, it is justified to remove or reduce the factor of 4 when determining the allowable throughwall circumferential flaw length. Again, this is justified since there was remaining ligament spread around the entire shroud circumference. If this is done, the observed indications, and resulting effective flaw lengths are well below the sllowable throughwall circumferential flaw lengths. It should be noted that the limiting allowable throughwall circumferential flaw is now governed by the L.EFM method and not the limit load method.

                                                                                                           ,,e a 49

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          ~

I l GENuclear Ewxy GENE-523-141-!M3, Rev. ) 5.6 References 5-1. Rooke, D.P. and Cartwright, D.J., " Compendium of Stress Intensity Factors," The Hillingdon Press (1976). 5-2. Ranganath, S., Mehta, H.S. and Nonis, D.M., " Structural Evaluation of Flaws in Power Plant Piping," ASME PVP Volume No. 94 (1984). l l e 0 e e

• 50

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s GENE-523141 lH3, Rev.1 GENuclear Energy f . Shroud Head Flange H1 i s'

                                                                                                  "                                     )

I Top Guide Support Ring ////) . , H3

                                                                              <        ,                 H4
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Core Plate Support Ring 7(( MA 7 H6 I

                                                                                           '                 H7 H8
                                                                              \                                            ,

Shroud Support Plate

                                                 '////////////////

RPV Shroud Support Cylinder Figure 5-1 51 , 4

                                                  ~

DELTR a (In.)

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DELTH a (mm)

 '                Figure 5-2 Comparison of J-R curves developed for two irradiated stainless steel specimens
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l GENuclear Energy _ GENE.523-141 1093, Rev. ! l . n A' I a-m , _ o/b F 0.0 1.0 3 0.1 1.006 4 2o

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! l l 1 Figure 5 3 SCHEMATIC ILLUSTRATING FLAW INTERACTIO I l 53

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9 M N I t , l 55 .

GENuclear Energy GENE.323141 1093, Rev.1 IW1 Y Mop All Flow Indcotions , lf Proximity Rules ' Map Effective Flow Lengths Circumferentic (Figure 2-4) " Flows l if if Axial Flows

                                                                      /  Sum of Flows in any 90 0 No i

l Sector <107.5" lf Yes No

                             <0.75(Lleft +Qff                                   II for Adjocent Flows?

y No S<0.75(Li erf +L2 eft For Mjacent flows? L= L1 egg or L2egr Yes I I ye3 l No Yes L= Llegg + derf ~

                                                                      "Ueff        CII U

No Fudher b 1.< 59"7 > Evoluotion ' / s Yes 3f Continued Operation Justified 4 Figure 5-6 SCHEMATIC OF SCREENING CRITERIA 56

l a GENuclear Energy GENE-523-1411093, Rev.1 6.0

SUMMARY

AND CONCLUSIONS An evaluation of the indications in the Peach Bottom Unit-3 core shroud has been performed to demonstrate that the structural integrity of the shroud is assured for the next two year cycle. In addition, the report documents material, water chemistry and fluence information which provides additional insight to the shroud condition. The primary focus of this report was to demonstrate that even with several conservatisms l in the evaluation, the stmetural integrity of the shroud is maintained during a limiting l event. This was performed by developing a screening criteria, assuming throughwall indications, which can determine the acceptability of the flaws based solely on the IVVI results. The assumption of through wallindications removes any uncertainty regarding sizing and the need to further characterize the indications. By meeting the screening criteria, the ASME Code Section XI safety margins are satisfied. The screening criteria uses both linear elastic fracture mechanics (LEFM) and limit load j concepts to determine acceptable through-wall indica, tion lengths. The limiting flaw length based on either LEFM or limit load was used for the screening criteria. l The screening criteria also uses the ASME Code Section XI criteria for combining flaws . based on the proximity ofindications. In addition, a second method for including the interaction between neighboring indication tips was considered for the LEFM allowable flaw size calculation. The resulting effective flaw lengths were compared against the screening criteria to determine if the structural integrity of the shroud was maintained. Based on the results of the application of the screening criteria to the observed indications, it is concluded that the structural integrity of the shroud is maintained for the next fuel cycle. All effective indication lengths were shown to be less than the allowable flaw size. O e 57

. GENuclear Eneru PEhTs533-141-1993, Rev.1 APPENDIX A DETERMINATION OF THE EFFECTIVE FLAW LENGTH The effective flaw lengths are based on ASME Code, Section XI proximity criteria as presented in Subarticle IWA-3300. The procedure addresses both circumferential and axial flaws. Indications are considered to be in the same plane if the perpendicular distance between the planes is less than 4" (2 times the shroud thickness). All flaws are considered to be through-wall. Therefore, indications on the inside and cutside surface should be treated as if they are on the same surface. When two indications are close to each other, rules are established to combine them based on proximity. 'Ihese rules are described here. A.1 Proximity Rules The flaw combination methodology used here is similar to the ASME Code, Section XI proximity rules concerning neighboring indications. Under the rules, if two surface indications are in the same plane (perpendicular distance between flaw' planes <4") and are within two times the depth of the deepest indication, then the two indications must be considered as one indication. In Figure A-1, two adjacent flaws L1 and L2 are separated by a ligament S. Crack growth would cause the tips to be closer. Assuming a conservative crack growth rate of 5x10-5 in/hr, crack extension at each tip is 0.8 in. for 16,000 hours or one fuel cycle (See Appendix B for crack growth rate discussion). Therefore, combining the crack growth and proximity criteria, the flaws are assumed to be close enough to be considered as one continuous flaw if the ligament is less than (2 x 0.8 + 2 x shroud thickness). For a shroud thickness of 2.0 in., this bounding ligament is 5.6 in. Thus, if the ligament is less than 5.6 inches, the effective length is (Ll+L2+S+1.6"). Note that the addition of 1.6 in. is to include crack growth at the other (non-adjacent) end of each flaw (See Figure A-2). If the ligament is greater than 5.6 in., then the effective flaw length is determined by adding the projected tip growth to each end of the flaw. For this example, Lleg= L1 + 1.6", and L2 eft = L2 + 1.6". A similar approach is used to combine flaws when a circumferential flaw is close to an axia! flaw (See Figure A-3). If the ligament between the flaws is less than 4.8 inches, then A-1

GENuclear Enerv GENE.5831411093, Rev. l

        ~

I 2. the effective flaw length for the circumferential flaw is Leff = Ll+S+0.8" (the bounding  ; ligament for these cases). If the ligament is greater than 4.8 in., then the flaws are treated l separately. l Mer the circumferential and axial flaws have been combined per the above criteria, a map 4 cithe effective flaws in the shroud can be made, and the effective flaw length can be used for subsequent fracture mechanics analysis. .i In order to demonstrate the proximity criteria, three examples are shown in Table A-1 and described below. i ' J l I Table A-1 Flaw Combinations Considered in Proximity Criteria l Case Circumferential Flaw Axial Flaw A Yes No B Yes Yes C No Yes s A.1.1 Case A: Circumferential Flaw - No Axial Crack , This case applies when two circumferentialindications are considered. Figure A-2a shows this condition. If the distance between the two surface flaw tips is less than 5.6", the indications must be combined such that the effective length is (See Figure A-2b): L efr= Ll + S + L2 + 1.6" where: L1 = length of first circumferentialindication L2 = length of second circumferential indication S = distance between two indications O 4 d A-2

l GENaclear EnerxY GENE-583141-1993, Rev.1 If the distance between the two tips is greater than 5.6", the effective flaw lengths are (See Figure A-2c): Lleg= L1 + 1.6" L2eg= L2 + 1,6" l A.I.2 Case B: Circumferential Flaw - Axial Flaw This case applies when both a circumferential and an axial flaw are being considered. Figure A-3a demonstrates this condition. For this case, only. growth of the circumferential flaw is considered. If the distance between the circumferential indication tip and the axial indication is less than 4.8", then the effective circumferential flaw length is (See Figure A-3b): L efr= L1 + S + 0.8" where: L1 = length of circumferential indication l S= distance between the circumferential tip and axial flaw. ' l and the effective axial length is (Figure A-3b): L eg = L2 + 1.6" . I where: L2 = length of axialindication j 1 If the distance between the circumferential indication tip to the axial indication is greater than 4.8", then the flaws are not combined (See Figure A-3c) and the effective lengths are: L1egr= L1 + 1.6" (for circumferential flaw) L2efr= L2 + 1.6" (for axial flaw) A.I.3 Case C: No Circumferential Flaw - Axial Flaw This case applies to when only axial flaws are being considered. The effective length is determined in a manner similar to that usedfor case A for circumferential flaws. 1 A-3 i

i l . GENuclear Energ GENE 533141-1093, Rev.1 A.2 Application of Effective Flaw Length Criteria The application of the effective length criteria is applied to two adjacent indications at a time. Figure A-4 is a schematic which illustrates the process. For example, using the 0* azimuth as the starting location for a circumferential weld or plane, the general procedure would be as follows:

     . Moving in the positive azimuthal direction, the first indication encountered is indication 1.
     . The next indication is indication 2.
     . Apply proximity .ules to the pair ofindications (indications 1 and 2). Combine the flaws if necessary (Ll+L2+S). Old indication 2 becomes new indication 1.
     . Continue along positive azimuthal direction until the next indication is encountered. This becomes new indication 2.
     . Apply proximity rules to new indications 1 and 2.
     . Continue proximity rule evaluation until all indications along the subject weld or plane have been considered.

d q l

                                                                                          ..        j A4 1
                                                                                                    )

I e GENuclear Eurgy GENE-3231411093, Rev !  ; t , 1 4 > x s h JL l i l 1 i I , I l I l Combined L1 raw i l l Flaw i . 1 l 1 , I l 1 i U i l D1 i L 1 I S 1 D2  ! U < > . l JL L2 re 2 l

                                                   <                                    l-i 1

1 I I I U U l l . Figure A 1 - ASME Code Proximity Criteria.. t l A-5

l GENuclear Energy GENE-523141 1093, Rev. ! l . I I no.s usumee mrougn- an l l l

                                                                    --      '                As-Found (2a)       _"                                                         \

4 L1  ? 4 s- = 4 L2 + na.. u.um.o mrougn-.e S < 5.6"

                                                                                      ';    'ef f -ti +s +t2 + s .e-(2b)      w.io 4                          L                       =                                                         ,

l l l T no. u.um.o mrougn-.e . S> 5.6"

                                                                                               '1eff "'1 + 1.6" (2c)      w.e                                                            T L2,gg =L2+ 1.6" (L1      ? C               s      ?          4   L2 - >

Figure A APPLICATION OF PROXIMITY PROCEDURE TO NEIGHBORING CIRCUMFERENTIAL FLAWS . A6 i l

1 l GENuclear Energy GEVE-523141-1093, Rev. I l g Flows Assamed Throughwoll n u A As-Found weio 7, 3 (30) - Y l l

                                                                                                )

l

                     + L1-      # <-S          >

( l l d T Flows Assumed Throughwoll

                                                            "           S < 4.8" weio
                                                   /                   L1eff =L1 +S+0.8'

. (3b) L2eff =L2+ 1.6"

                                                  )     if l
                      + L1-      ?    4 -- s    >                                               I l

4 L = f T Flows Assumed Throughwoll h L2 S > 4.8"

                                                                    \     U        L1 + 1.6" (3c)-      weio i                T   L2 eff =L2+ 1.6" eff =
                                                     )   il 45- Lt    >    4     $ -+                                           .

i Figure A- APPLICATION OF PROXIMITY PROCEDURE TO NEIGHBORING AXlAL AND CIRCUMFERENTIAL FLAWS .. A-7 .

4 GENuclear Enerv GENE-523-141-1093, Rev. ! Stort at Theto=0 Move in + Theto Direction lf i=1 l II First Flow is Flow I lI

                               ^

Next Flow is Flow i+1 If Perform Effective Length Calculation lf Combine Flows If Necessary To Determine Effective Length 1=i+1 ]f j( Flow i+1 = Flow I lf No Lost Flow? Yes

• Done Figure A PROCESS FOR DETERMINING EFFECTIVE CIRCUMFERENTIAL FLAW LENGTH A-8 ,

 - -       - _      _      __ -_         ___..     -- -- . -               __ - - - - - ~_ -_                           _ -_-.
     -         GENulm EmW                                                                     GENE-523-141 !M3, Rev.1 l

APPENDIX B BASIS FOR THE CRACK GROWTH RATE The basis for the crack growth rate used in the screening enteria is provided in this section. The Peach Bottom Unit 3 shroud cylinder was fabricated from roll formed Type l 304 stainless steel plate. Therefore, the weld heat-affected-zone (HAZ) is likely l sensitized. The shroud is also subjected to neutron fluence during the reactor operation which further increases the effective degree of sensitization. The other side-effect of l neutron fluence induced irradiation is the relaxation of weld residual stresses. The slip-dissolution model developed by GE quantitatively considers the degree of sensitization, the stress state and the water environment parameters, in predicting a stress corrosion cracking (SCC) growth rate. The crack growth rate predictions of this model have shown good correlation with laboratory and field measured values. This model was used to predict a Peach Bottom Unit-3 specific crack growth rate and a conservative value was then selected based on this value. B.1 Slip-Dissolution Model Figure B-1 schematically shows the GE slip-dissolution film-rupture model (Reference B-1) for crack propagation. The crack propagation rate V tis defined as a function of two ) constants (A and n) and the crack tip strain rate. The constants are dependent on material I and environmental conditions. The crack tip strain rate is formulated in terms of stress,  ! loading frequency, etc. When a radiation field, such as the case for the shroud, is present, there is additional interaction between the gamma field and the fundamental parameters which affect intergranular stress corrosion cracking (IGSCC) of Type 304 stainless steel (see Figures B-2 and B-3). The increase in sensitization (i.e., Electrochemical Potentiokinetic Reactivation, EPR) and the changes in the value of constant A as a function of neutron fluence (>lMev) is given as the following: EPR = EPRo + 3.36x10-24 (fluence)l.17 (B-1) where, EPR is in units of C/cm2, fluence is in units of n/cm2and the calculated value of , EPR has an upper limit of 30. B-1

   - ..   . .      .-                                       .                       ~                      . _   -
        .                                                                                                              l l

l GENuclear Enerv GENE-383-141-1093, Rev. I The constant A is defined as the following: l ! for fluence 5 1.4x1019 n/cm2: C = 4.1x10-14 (B-2a)

for fluence > 1.4x10 19 n/cm2 but 5 3x1021 n/cm2

, (B-2b) '

l C = 1.14x10-13 In(fluence)-4.98x10-12 for fluence 51.4x1019 n/cm2: C = 4.1x10-14 (B-2c) The units of K to be used with the above expressions is MPaVm. B.2 Calculation ofParameters The parameters needed for the crack growth calculation are: stress state and stress intensity factor, effective EPR, water conductivity, and electro-chemical corrosion potential (ECP). The stress state relevant to IGSCC grow s rate is the steady state stress which consists of weld residual stress and the steady applied stress. Figure B-4 shows observed through-wall weld residual stress distributions for large diameter pipes. This distribution is expected to be representative for the shroud welds also. The maximum stress at the surface was nominally assumed as 35 ksi. The steady applied stress on the shroud is due to core differential pressure and its magnitude is small compared to the weld residual stress magnitude. Figure B-5 shows the assumed total stress profile used in the evaluation. Figure B-6 shows the calculated values of stress intensity factor (K) assundng a 360' circumferential crack. It is seen that the calculated value of K reach'6s a maximum of approx. 25 ksiVin. The average value of K was estimated as 20 ksiVin and was used in the crack growth rate calculations. The weld residual stress magnitude is expected to decrease as a result of relaxation produced by irradiation-induced creep. Figure B-7 shows the stress relaxation behavior of Type 304 stainless steel due to irradiation at 550' F. Since most of the steady state stress in the shroud comes from the weld residual stress, it was assumed that the K values shown in Figure B-6 decrease in the same proportion as indicated by the stress relaxation , behavior of Figure B-7. B2

  . . - . . . - . _ - _          -   .   . . -      - _ - - -    . - - - _ - - ~ .            . . - . _ . _ . .              .     - - - - -

i

• GENucle:r Energy GENE-523-141 1093, Rev. !

The second parameter needed in the evaluation is the EPR. In the model, the initial EPR . value is assumed as 15 for the weld sensitized condition. Using Equation (B-1), the predicted increase in EPR value as a function of fluence is shown in Figure B-8. The third parameter used in the GE predictive modelis the water conductivity. The  ; reactor water conductivity at Peach Bottom Unit-3 has recently (1992-93) been good I (approx. 0.1 pS/cm2) compared to earlier operating period (1975-84) when the average conductivity was in excess of 0.7 S/cm .2This has a significant impact on the predicted ! crack growth rate by the GE model as seen in Figure B-9. To demonstrate that the GE model conservatively reflects the effect of conductivity, Figure B-10 shows a comparison l of the GE model predictions with the measured crack growth rates in the crack advance verification system (CAVS) units installed at several BWRs. The comparison with CAVS data in Figure B-10 also demonstrates the conservative nature of crack growth predictions l

by the GE model. l 1

! The last parameter needed in the GE prediction model is the ECP. Figure B-11 shows the j measured values of ECP at two locations in the core. Since Peach Bottom Unit-3 does l not plan to use hydrogen injection, the ECP values at zero H2 injection are relevant in Figure B-ll. It is seen that the ECP values at zero H 2injection rate range from 150 mV l to 225 mV. Therefore, a value of 200 mV was used in the calculation. e B.3 Crack Growth Prediction Based on the discussion in the preceding section, the crack growth rate calculations were conducted as a function of fluence assuming the following values of parameters: Initial K = 20 ksidin EPRo = 15 C/cm2 Cond. = 0.1 pS/cm2 ECP = 200 mV i Figure B-12 shows the predicted crack growth rate as a function of fluence. It is seen that, the predicted crack growth rate initially increases with the fluence value but decreases later as a result of significant reduction in tiie K value due to irradiation induced stress relaxation. The crack growth rate peaks at 4.5x10-5 m/hr at a fluence of lx1020 n/cm2, The crack growth rate for the expected fluence at the end of the next operating cycle is B-3

GENulcar FaerKY GENE-523-1411M3, Ren 1 i approximately 2x10-5 n/hr. Thus, a bounding value of 5x10-5 n/hr i can be conservatively used in the structural integrity evaluations for the shroud. This bounding crack growth rate is quite conservative as can be shown in Figure B-13 from NUREG-0313, Rev. 2. It is seen that the crack growth rate of 5x10-5 in/hr at 20 ksiVin is considerably higher than what would be predicted by using the NRC curve. This further demonstrates the conservatism inherent in the assumed bounding value of crack growth rate. B.4 Summary A crack growth rate calculation using the GE predictive model was conducted considering the steady state stress, EPR, conductivity and ECP values for the Peach Bottom Unit-3 shroud. The evaluation accounted for the effects ofirradiation induced stress relaxation and the increase in etE-etive EPR. The evaluation showed that a bounding crack growth rate of 5x10-5 n/hr i may be conservatively used in the structuralintegrity evaluation of the Peach Bottom Unit-3 shroud. B.5 Reference B-1 F.P. Ford et al, " Prediction and Control of Stress Corrosion Cracking in the Sensitized Stainless Steel / Water System," paper 352 presented at Corrosion 85, Boston, MA, NACE, March 1985. IP O e sa

• B-4

                                                                                                    )

GENE-585-141-1093, Rev.1 GENuclear Energy

    ~

The GE PLEDGE Slip Dissolution - l Film Rupture Model of Crack Propagation l 1 ! E cT '

                                 ~

a l l N crack-tip advance by l N VT enhanced oxidation at i strained crack tip  ! l I i !, .n ! VT = AE CT i I Where: VT - rack propagation rate i A, n = constants, dependent on material ' and environmental conditions 4 l E CT crack-tip

                         -           st. rain rate, formulated in terms of stress, loading frequency, etc.

l 4 Figure B-1 B-5 ,

                                                                                                                                                                                                                                   . l
                                                                                                                                                                                                                           . .       t l

t SOLUTION RENEWAL RATE TO CRACK-TIP si ' N STRESS y ii . A$ ANIONIC \ OXIDE RUPTURE gEl

                ,   TRANSPORT                                                                                                                    RATE AT ENVIRON-                                                                                             CRACK-TIP MICRO-y-FIELD                                                                                                                STRUCTURE l

! J WAN CRACK TIP $(A)~, pH e IRSSIVATION RATE

    '                                    AT CRACK-TIP                                                                                   :                                N-FLUENCE                                                   l GS.DENUDATW                          "d .ZGATION
                                                                                                                                       '                                                                                        E i                                                                                                                                                                                                                               5 5
                                                                                                                                                                                                                                $l i        Figure B-2 Effects of Fast Fluence, flux & Gamma Field on parameters affecting IGSCC of Type 304 Stainless Steel 1  -

i n VT=AE cf Q iF ID i' 0 Jl ( .

                                                                                                                  /

M+ K

                                             =

Ja , JH20 g r > E et r1 u c JM+ U ( '

    ?                                                                                                         t      Cr       R S, Ni, Si hc                                                                                       h
                                                                .                                               n Vy = 7.8 X.tO-3         n3.6 g x ig-14 g 4                                                                                                                        :

f(K) ' f(EPR) . o '

                                                                 -                                                                                                                    D n=<

u ef(K) +e f(6)cs

                                                                                                          -                                                                           t E

5

                                                                                                                                                                                      =         ,

9

• m -

2

• I Figure B-4 Parameters of FundamentalImportance to Slip Dissolution Mechanism of IGSCC in Sensitized Austenitic Stainless Steel r

  -                                                                                                                                                                                             t

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d GENE PLEDGE Model Prediction for PB-3 a Sensitized Type 304 Crack Growth Rate [D Crack Growth Rate, in/h . . . - . . . . - _ . _ 200 mV 5.000E-05

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GL 94-03 ! g v I PECO ENERGY d magene,, 965 Chesterbrook Boulevard ( Wayne. PA 19087 5691 November 7,1994 1 Docket No. 50-277 License No. DPR-44 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Subject:

Peach Bottom Atomic Power Station, Unit 2 Supplemental Response to Generic Letter 94-03 i Summary of Core Shroud Inspection Results '

Dear Sir:

l in our letters from G. A. Hunger, Jr. (PECO Energy Company) to U. S. Nuclear Regulatory l Commission (USNRC), dated August 24,1994, September 9,1994 and September 26,1994, t PECO Energy Company provided inspection plans for the Peach Bottom Atomic Power Station (PBAPS), Unit 2 core shroud. These plans were submitted in accordance with Reporting Requirements 1 and 2 of Generic Letter (GL) 94-03, 'Intergranular Stress Corrosion Cracking of l Core Shrouds'in Boiling Water Reactors.' A preliminary summary of the inspection results and a i preliminary evaluation of the results were provided in a letter dated October 17,1994. I The purpose of this letter is to provide the final summary report, as requested by Reporting  ; l Requirement 3, of the GL - I i in summary, the overall results of the inspection revealed a minimum amount of flaws. Less , j than 5% of the examined weld length was found to contain flaws. The evaluation of the results l j was performed following the approach outlined in the 'BWR Core Shroud Inspection and Flaw i Evaluation Guidelines,' GENE-523113-8094, dated September 1994. This evaluation, based on l the examination data, concludes that there is a substantial margin for each of these welds under  ! conservative, bounding conditions to allow for continued operation of PBAPS, Unit 2. If you have any questions, please contact us. t Very truly yours, e e 1. G. A. Hunger, Jr., Director - Licensing Attachment, Affidavit cc: T. T. Martin, Administrator, Region 1, USNRC W. L Schmidt, USNRC Senior Resident inspector, PBAPS

l November 7,1994 *

      . Page 2 bec:     R. A. Bur.icelli, Public Service Electric & Gas

• R. R. Janati, Commormealth of Pennsylvania R.1. McLean, State of Maryland j H. C. Schwemm, Atlantic Electric A. F. Kirby, Ill, Delmarva Power & Light Company D. M. Smith - 63C-3 t

G. R. Rainey - PB, SMB4-9 l W. H. Smith, lli 62C-3 l J. B. Cotton 51 A-1  ; J. Doering - 63C-5 l ,. G. D. Edwards - PB, A41S F. W. Polaski/ISEG - PB, SMB4-6 M. C. Kray/TRL - 62A 1 l A. J. Wasong/DJF - PB, A4-SS ! C. J. McDermott - MO, S131 J. T. Robb 610-1 D. L Schmidt LGS, SSB2-3 J. J. Stanley PB, SMB34 , R. E. Ciemiewicz 638 3 l M. E. Kowalski - 63B-3 N. Silvestri - 63B 3 i D. B. Fetters 63B-3 V. M. Nilekani - 63C-9 T. A. Moore - PB, SMB34 T. J. Niessen - SMB3-2 Commitment Coordinator - 62A-1 Correspondence Control Desk - 61B4 DAC 61B-5 ds\planrest l L l l l

   - -.        .                    .      - . _           - - .       .       -..-... .                .- ~. .        - . -

I PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION .t . UNW 2 i REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS / FINAL REPORT 2R10, October 1994 Docket No. 50 277 In September and October of 1994, during the tenth refueling outage of Peach Bottom Atomic Power Station (PBAPS), Unit 2, the core shroud structure was comprehensively inspected. These inspections were conducted to determine the condition of the shroud welds, relative to the potential for existence of Intergranular Stross Corroslor? Cracking (IGSCC). The effort satisfied the commitment made for PBAPS, Unit 2, in the PECO Energy responses to NRC Generic Letter 94-03, included in PECO Energy letters dated 8/24/94, 9/9/94, and 9/26/94. The inspections were , conducted in accordance with the guidance provided by the Boiling Water Reactor Vessel and ' Internals Project (BWR VIP), as presented in the "BWR Core Shroud Inspection and Flaw Evaluation Guidelines", GENE 523-113-0894, dated September 1994 (Reference 1). The following describes the overallinspection effort and summarizes the results of this effort. Further, additional plans to assure continued core shroud integrity are presented. BACKGROUND: The PBAPS, Unit 2 shroud was fabricated by Rotterdam Drydock Co. LTD., Rotterdam, Holland. The product forms used for this fabrication included 2" thick ASTM A240, Type 304 stainless steel plate (for shroud cylinders), and ASTM A182, Grade F304 seamless, stainless steel rolled forgings (rings). The plate materials contain relatively high carbon contents (.059% to .062%), while the ring forgings contain lower carbon contents (.028% to .035%). The product forms where joined using the submerged arc welding process.The weld filler metal used was ASTM A371 Type Er308, with low carbon content. Welds H-1 through H-6 were welded from both surfaces, using a double bevel weld prep. Weld H 7 was welded from the inside surface of the shroud using a single bevel weld prep and a backing ring. The H 7 weld was made at the PBAPS site, and it attached the pref abricated shroud structure to the Reactor Pressure Vessel. This weld is a dissimilar metal weld (304 stainless to Alloy 600). The filler metal used for this weld was ASTM B 304, Type ERNiCr-3 (Alloy 82). The process used for this joint was the Shielded Metal Arc Welding process. Attachment 1 includes a drawing which depicts the shroud configuration, weld locations, and materials of fabrication. The PBAPS, Unit 2 shroud has been in service since July 1974. During the first decade of hot operation, PBAPS, Unit 2 operated with relatively high primary water conductivity. Unit 2's arithmetic mean conductivity actually exceeded 1.0 pS/cm during the first few years of operation. Subsequently, conductivity values were steadily decreased to below current EPRI guidelines.1992 and 1993 values were actually less than 0.1 pS/cm. The effects of such early water chemistry history on the susceptibility of the shroud w' elds to IGSCC are addressed in Reference 1. The above described factors place the PBAPS, Unit 2 shroud into inspection Category C, as defined by Reference 1. This category has a high potential for some amount of shroud cracking, and, therefore, comprehensive inspections of welds H-1 through H-7 is recommended. Page 1 of 4 L

I COMMONWEALTH OF PENNSYLVANIA  :

ss.

l COUNTY OF CHESTER  : l

 ~

W. H. Smith, ll!, being first duly swom, deposes and says: That he is Vice President of PECO Energy Company; that he has read the enclosed response to Generic Letter 94-03 dated Jtly 25,1994, for Peach Bottom Facility Operating License DPR44 and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, Information and belief. l V Vice President l Subscribed and swom to before me this day of 1994. h  % Nota // ry Public Natanalsees

             %%Jas          CD*

PECO ENERGY COMPANY l PEACH BOTTOM ATOMIC POWER STATION UNIT 2 REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT 2R10, October 1994 Docket No. 50-277 INSPECTIONS: The scope of the core shroud inspections included all of the shroud circumferentialwelds (e.g. H-1 through H 7). The method used forinspection of these circumferentialwelds was Ultrasonic Testing (UT), performed from the outside surface of the shroud, using the General Electric Company's SMART 2000 data acquisition system and the General Electric Company's OD Tracker and Suction Cup Scanners. The extent of the planned inspections included all portions of the circumferential welds which were accessible for the above described equipment. This scope and extent of planned inspections was identified in PECO Energy's second response to Generic Letter 94-03, dated September 9,1994, and clariflad during subsequent communications with the NRC. The UT scanning was accomplished using three transducers. These transducers included 45' shear wave, 60* longitudinal wave, and creeping wave units. The transducers scanned 6ach Heat Affected Zone (HAZ) of the accessible lengths of each weld. The creeping wave transducer was not used on the H-3 weld inspection, due to equipment f ailure. The creeping wave transducer was used to enable better near-surface detection capabilities, primarily on the ring side HAZ of welds H-1, H 2, H 5, and H-6. On the plate side HAZ of the welds, the 45' shear wave transducer typically was able to achieve a full V sound transmission, enabling near-surface detection with this transducer. l The purpose of the shroud inspections was to assess the condition of the shroud circumferential I welds so that the integrity of the shroud structure could be quantitatively or qualitatively l demonstrated. Additionally, the inspection results will be used to establish a baseline of this , condition for comparison to future inspection results.This baseline data and subsequent inspection l results will also be used to develop schedules for future shroud inspections, evaluations, or repairs. l The extent of shroud weld inspections performed during 2R10 include: l 33% of the length of Weld H 1, distributed over 66% of the weld circumference. 230" 84% of the length of Weld H-2, 583" 88% of the length of Weld H-3, 574" 89% of the length of Weld H-4, 580" 83% of the length of Weld H-5, 540"

           .10% of the length of Weld H-6, plus an additional 13% performed visually.                    143" 09% of the length of Weld H 7.                                                                59" Subtotal                 2714" x 2 (HAZ per weld) l Total                    5428"     l The extent of these weld inspections is graphically depicted on the attached weld maps,                i Figures 1.1 through 1.7 (Attachment 2).

Page 2 of 4

 - - . - - . - - - - ~ .                                    . _ - - . _ - . - -                    - - - - - - , - . - -.                  .-

PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION , UNW 2 REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT 2R10, October 1994 Docket No. 50-277 l RESULTS: A sufficient length of welds H-2, H-3, H-4, and H 5 was inspected to quantifiably demonstrate the condition and, therefore, the structural integrity of these welds. For welds H 1, H-6, and H 7, a , limited amount of weld length was inspected, due to accessibility for inspection equipment. In these ' cases, the welds received a qualitative evaluation of their structural integrity. This qualitative approach is identified and discussed in Reference 1. Some indications were found on welds H 1, H-3, H-4, H 5, and H-6. No indications were found on welds H 2 and H 7. Tables 1.1 through 1.7 provide data on the as-found indications, and are included as Attachment 3. Because the H-6 weld had limited UT inspection, both in extent and distribution, and because one indication was identified, an additional scope of inspections was implemented for this weld. This additional scope included an enhanced visual inspection (VT-1) of all remaining accessible areas of this weld from the OD. These areas included the spaces between the ten sets of jet pumps. The weld was cleaned before the inspections. l 1 EVALUATIONS: 1 For welds which were accessible for a comprehensive inspection, all as-found indications were , assumed to be through wall. Therefore, depth sizing of the indications was not utilized. Additionally, the weld lengths which were not inspected, due to inaccessibility, were also assumed to be through wall indications. / For welds which had limited accessibility, the conditions found within the inspected weld length were extrapolated over the areas of the weld which were inaccessible for inspection. This extrapolated condition was then evaluated for structural integrity. For welds H 1, and H-6, the l evaluation assumed that the entire circumference of the weld had an indication equal in size (i.e. depth) to the deepest Indication actually found on the weld by the inspections. The remaining safety margin was then calculated. Since no indications were found on weld H-7, calculations were developed to determine the maximum initial indication size (assuming a 360' length), which could

                              ' .be tolerated. This information was then used for qualitative comparison to the as-found conditions.

inspection results, for welds which received a comprehensive inspection, were initially compared ! to a screening criteria, which had been developed prior to the inspections, if the results of the inspection Indicated that sufficient unflawed material existed in the correct locations around the l weld circumference, the results were considered acceptable. For welds which received a limited inspection, and for comprehensive inspection results which did not meet the original screening t criteria, a detailed evaluation was performed. Ultimately, a detailed evaluation was performed for all welds, to determine the margin of safety for each weld (see Table 3 and 4 in Attachment 4). l The detailed evaluations were performed by Structural Integrity Associates (SI). These evaluations ! used the guidance provided in the evaluation portion of Reference 1. The as-found indication l lengths were adjusted for upper boumi crack growth, NDE uncertainty, and proximity f actors. The Page 3 of 4

PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION UNIT 2 REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT 2R10, October 1994 Docket No. 50-277 resultant indication lengths (as evaluated indications) were then used to calculate the amount of safety margin remaining in the subject weld, using the limit load methodology. Additionally, for Welds H 3 and H-4, the Line ar Elastic Fracture Mechanics (LEFM) technique was used, due to the extent of neutron exposure received at these weld locations. The safety factors was calculated against the most limiting design basis loading conditions, dorived from the General Eledric Nuclear Energy Screening Criteria Document (Reference 2) and the PBAPS, Unit 2 UFSAR. The loadings also considered Power Rerate conditions. The as-evaluated indication lengths are graphically depicted on the attached weld maps, Figures 1.1 through 1.7, (Attachment 2). A more detailed discussion of the evaluations, including factors utilized for crack growth and NDE uncertainties, is contained in the Structural Integrity Associates' final report No. SIR-94-111, Rev.0, included as Attachment 4. CONCLUSIONS: A 10CFR50.59 determination and safety evaluation has been developed and reviewed by the Plant Operat!ons Review Committee (PORC). The conclusion of this evaluation indicates that no unreviewed safety questions exist as a result of the shroud inspection findings. The results of the inspections and evaluations conclude that the condition of the PBAPS, Unit 2 shroud, projected through the next, two-year operating cycle, will support the required safety margins, specified in the ASME Code and reinforced by the BWR-VIP recommendations. Additionally, the results of these UT inspections substantiate the assumptions made for Unit 2, during the PBAPS, Unit 3 shroud visual inspections, and the Safety Analysis developed in response to Generic Letter 94-03. The extent of the shroud inspections provide a comprehensive baseline for comparison to future inspections. In accordance with Reference 1, welds H-1, H-6, and H-7 require reevaluation and/or reexamination during the next refueling outage (2R11). Welds H-2, H 3, H-4, and H-5 require reevaluation and/or reexamination during the following refueling outage (2R12). The results of the UT inspection of the two access hole cover welds revealed no indications. PECO Energy will continue to follow the developments of the BWR-VIP guidance documents, and will evaluate their applicability to the PBAPS Site.

REFERENCES:

l 1. BWR Core Shroud Inspection and Flaw Evaluation Guidelines, GENE-523-113-0894, September, l 1994.

2. Evaluation and Screening Criteria for the Peach Bottom Unit 2 Shroud, GENE-523-176-1293, i December 13,1993.

Page 4 of 4

l l r l l l PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION l UNIT 2 REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT 2R10, October 1994 Docket No. 50-277 ATTACHMENT 1 l l

PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION l UNIT 2 l REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT 2R10, October 1994 Docket No. 50-277 l ATTACHMENT 1 REACTOR PRESSURE VESSEL - SHROUD PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 p DRYER / SEPARATOR SUPPORT RING " WELO NO. \ M ITEM 7 A182-F 304 0.035 % C Hi jf V1 V2 g ITEM 5 A240 TP.304 0.062 %, C (MAX) H2%$ /-T,0P GUIDE SUPPORT RING, R%%1 ITEM 6 A182-F 304 0.028 % C (UNIT 2) H3 l' ITEM 4 A240 TP. 304l 0.030  % C (UNIT 3) 0.060 % C (MAX)

                       $    V3                         V4 H4                        V5                         V6                    /

ITEM 14 A240 TP. 304 0.060 % C (MAX) l H5 I

                      /

gCORE PLATE SUPPORT RING V/////J ITEM 13 A182-F 304 0.030 % C (UNI" 2) 0.035 % C (UNIT 3) l l ITEM 12 A240 TP.304 0.059 % C (MAX) H7 V7 V8 9 j SHROUD SU R YLINDER pswwwxj W- REF. DWG. M-1-B-26

l l l PECC ENERGY COMPANY

 ,         PEACH BOTTOM ATOMIC POWER STATION UNIT 2 REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT 2R10, October 1994 Docket No. 50 277 i

ATTACHMENT 2 l l I l l l l l i l

2R10 LEGEND: PBAPS UNIT 2 i i = exAu AhEA CORE SHROUD INSPECTION =

UNINSPECTED WELD MAP FOR WELD No. H -1 4.12* EX AW ARE A FI GU RE NO.1.1 CENTERED BETWEEN (0 0* *

[g',2TE ED BET EN ' LUG SETS # 3 q LUG SETS Q 5 i F g gND CATIONS gy CATI 0ut {gt I g CATIO , A -E gCATIS 270* 5 E3 0 GREES. 5 SQ* 270* E I 1{ 8"OL]I 8- '

                                             =                                                                                                                                                                                                                                                            5 - 90
                                -286*                                                                                                    .

E 268* =- E E= -

                                                                                                                                                                                                              =                                                                                            -

W Y=A r_ N'_ _N 120* OF ~- ' I20* OF ~~ ~

                                                                                                                                                                                                                                                                                 /

UN ANTICI PATED ">, UNA NTICIPA TED Fj INTERFERENCE INTERFERENCE DUE TD (ISI LUGS 180 180*

• ON SHROUD 03.

DUE TO (IS) LUGS ON SHROUD GE. BETWEEN H-1 1 H-2 BETWEEN H-l & H-2 UPPER HEAT AFFECTED ZONE LOWER HEAT AFFECTED ZONE WELD LENGTH TOTAL INSPECTED PERCENT OF TOTAL WELD SHROUD CIRCUMFERENCE WELD LENGTH WELD LENGTH NUMBER 0.0.( I NCHES ) 0.D.( I NCHE S) (INCHES) INSPECTED ' H-1 UPPER 220" 691.2" 230" 33% H-1 LOWER 220" 891.2" 230" 33%

                                                                                                                                                                                                 '6 6"

• 2R10 .

PBAPS UNIT 2 '

                                                                                                                                                                                                                                           ' = {MM[ffo@RY CORE SHROUD INSPECTION                                                                                                                               -
                                                                                                                                                                                                                                             = UNINSPECTED WELD MAP FDR WELD No. H-2                                                                                                                                                                   m = pggF g          DNS

g. FI GURE NO.1.2 0' M = AS EVALUATED INDICATIONS

                                                                                                                                                                                         .-                                  ?
                        %. 5                                                                                                                                                      f'\. 5 NO INDICATIONS FOUND                                                                                                                                      '

NO INDICATIONS FOUND 270* 90* 270' 90* m - m 180* 180* UPPER HEAT AFFECTED ZONE LOWER HEAT AFFECTED ZONE WELD LENGTH TOTAL INSPECTED PERCENT OF TOTAL WELD SHROUD CIRCUMFERENCE WELD LENGTH WELD LENOTH NUMBER 0.0.( I NCH ES ) 0.0.( I NC HE S ) (INCHES) INSPECTED H-2 UPPER 220" 691.2" 583" 84% H-2 LOWER 220" 691.2" 583" 84%

LEGEND: 2R10 * = I N I C A TI O N. # . PBAPS UNIT 2 R = SATISFACTDRY CORE SHROUD INSPECTION INSPECTION

                                                                                                                                                              = UNINSPECTED WELD MAP FOR WELD No. H-3                                                                                      M = AS FOUND FI GURE         NO.1.3                                                                                                INDICATIONS O'                                                                                      0*                           M = AS EVALUATED INDICATIONS
              -                                                                                      1 9,__I*2 3
                                           .                                                                                                                                4 5
                                               '.                                                                                                                             6 270'       .                                           90'  270'                             .                                                                                             S O'
          .                          AL31                                            g.                                                                         SL31
                .                        .                                                               .                                                          .      9 Q
                                , -d '

15 I4 180* 180' UPPER HEAT AFFECTED ZONE LOWER HEAT AFFECTED ZDNE WELD LENGTH TOTAL INSPECTED PERCENT OF TOTAL WELD SHROUD CIRCUMFERENCE WELD LEN0TH WELO LENGTH NUMBER 0.D.( I NCH E S ) 0.D.( I N C H E S ) ( I NCHES) INSPECTED H-3 UPPER 207" 651" 574" 882 H-3 LOWER 207" 651" 574" 88%

LEGEND: 2RIO * = INDICATION- # PBAPS UNIT 2 ' CORE SHROUD INSPECTION '"5NSPEC 0

                                                                                                                                                                                                                                                                                                                           = uNINSPECTED WELD MAP FOR WELD No. H-4 FI GURE NO.1.4                                                     " " $50$0"8 IONS O'                                                            0*                    M = AS EVALUATED INDICATlDNS 1*                                           c 2

90' 270* S 0' 270* 3 ,. 5 2 2 6 g7 180* 180' UPPER HEAT AFFECTED ZONE LOWER HEAT AFFECTED ZONE WELD LENGTH TOTAL INSPECTED PERCENT OF TOTAL WELD SHROUD CIRCUMFERENCE WELD LENGTH WELD LENGTH NUMBER 0.0.( I NCH E S ) 0.0.( I NCHE S) ( I NCHES) INSPECTED H-4 UPPER 207" 851" 580" 89% H-4 LOWER 207" 651" 580" 89%  !

LEGEND: 2R10 * " "" PBAPS UNIT 2 , "= 33N EA3CT0RY T i CORE SHROUD INSPECTION -INSPECTION

                                                                                                                                                                         = UNINSPECTED WELD MAP FOR WELD No. H-5                                                                                                                  AS FOUND FI GURE                                      NO.1.5                                                          M = 1NDICATIONS
                            -                                                                                                                            M = AS EVALUATED 0                                                                                                           0                                   INDICATIONS t            #                                                                                  %                                   #

( 270* 90* 270' S 0*

                  =               -                                                                               =                                     /

180* 180' UPPER HEAT AFFECTED ZONE LOWER HEAT AFFECTED ZONE WELO LENGTH TOTAL INSPECTED PERCENT OF TOTAL WELO SHROUD CIRCUMFERENCE WELO LENGTH WELD LENGTH NUMBER 0.0.( I NC HE S ) 0.0.( I NC H E S) ( I NCHES) INSPECTED H-5 UPPER 207" 651" 540" 83% H-5 LOWER 207" 651" 540" 83%

LEGEND: 2R10 * = INDICATION #. PBAPS UNIT 2 I I = SATISFACTORY CORE SHROUD INSPECTION INSPECTION

                                                                                                                                                                                                                                                                                          = UNINSPECTED

• WELD MAP FOR WELD No. H-6 .

M = AS FOUND en FIGURE NO.l.G. INDICATIONS , O' 6 6* VT 0' h '

                                                                                                                                                                                                                                                                                    ~                    B* VT 1

315 ~~

                                                                                                                                               -'           1       -   1
                                                                                                                                                                                                   < 5-315               _A.          1' .                                          < 5-i OTE:                                                                                           OTE:

285 AS-EVALUATED 75 285 AS-EVALUATED 75  : INDICATION LENGTH INDICA TION LENGTH [ 270 g DNRES 90' 270 DNRE h 90'

                                                                                                                               ;                                                                    =

E t

                                                                                                                               =                                                                                                   t 255'                                                                                         105*  255'                                                                                            105' 225' 135 225'

• 135' 8 180' UPPER HEAT AFFECTED ZONE LOWER HEAT AFFECTED ZONE WELD LENGTH TOTAL INSPECTED PERCENT OF TOTAL WELO SHROUD CIRCUMFERENCE WELD LENGTH WELD LENGTH NUMBER 0.0.( I NC H E S ) 0.0.( I NC H E S ) (INCHES) INSPECTED <

H-6 UPPER 201" 631.5" G 4" U T. 84" YT 10% UT,13% VT  ! H-6 LOWER 201" 631.5" 64" UT 84" VT 10% UT.13% VT i

LEGEND

• 2R10 PBAPS UNIT 2 I I = SATISFACTORY CORE SHROUD INSPECTION "

                                                                                                                            !=    u !N!NCf"O WELD MAP FOR WELD No. H-7                                               M = AS FOUND
                                 $

• IN ICATIONS

                                               .           FIGURE N O . 1.7                                     .     .

h

• o,%wy>

s. .w' y' - -'

                                                                                                                       ~ ~ ~ ,,

g NO INDICATIONS FOUND 3 _ NO INDICATIONS FOUND g 270' s

                                                                   =

90' 270' s1. =

                                                                                                                                                            $         90' D_ -

x- / n yx m' -

                        .,                     : /                                             ym                      :,: ; .- -

O/ 180' "*. 3 1 8 0' UPPER HEAT AFFECTED ZONE LOWER HEAT AFFECTED ZONE WELO LENRTH TOTAL INSPECTED PERCENT OF TOTAL WELO SHROUD CIRCUMFERENCE WELO LENGTH WELD LENOTH NUMBER 0.0.( I N CHE S ) 0.0.( I NC H E S ) ( I NCHES) INSPECTED H-7 UPPER 201" 831.5" 59" 9% H-7 LOWER 201" G31.5" _ 59" 9%

I l PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION UNIT 2 l REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT l 2R10, October 1994 Docket No. 50 277 ATTACHMENT 3 l 1 l l l i 1 I l t t i

PECO ENERGY COMPANY - i PEACH BOTTOM ATOMIC POWER STATION UNW 2 - -

SUMMARY

OF CORE SHROUD EXAMINATION RESULTS r 2R10 TABLE 1.1 , Weld No. H1  ; Azimuth Location f Indication # Surface Weld Side Length

• Depth Remarks [

Start Stop [ t 1 ID Lower 12.87* 15.02* 4.13* .48' Detected by 45*S/60*RL [ 2 ID Lower 20.09* 21.40' 2.52* .61* Detected by 45*S/60*RL 3 ID Lower 27.81* 29.63* 3.50' .55" Detected by 45*S/60*RL 7 4 4 ID Lower 38.22* 38.33* 0.21" .31

• Detected by 45*S/60*RL l 5 ID Lower 42.59' 45.02* 4.67* .63* Detected by 45*S/60*RL 6 ID Lower 57.59' 59.52* 3.71* .54" Detected by 45*S/60*RL i

7 ID Lower 73.36' 75.83* 4.75* .74* Detected by 45*S/60*RL ' 1 8 ID Lower 80.92* 83.11* 4.21" .71* Detected by 45'S/60*RL 9 ID Lower 320.20* 321.41* 2.33* .54* Detected by 45'S/60*RL 10 ID Lower 327.59* 328.90* 2.52' .35* Detected by 45'S/60*RL - t i 11 ID Lower 330.11* 330.83* 1.38' .41* Detected by 45'S/60*RL j

• Indication Length - As developed on outside surface of shroud. (Using 1.923*/Deg.)

l L l t k

PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION UNIT 2 s . -

SUMMARY

OF CORE SHROUD EXAMINATION RESULTS 2R10 TABLE 1.2 Weld No. H2 Azimuth Location Indication # Surface Weki Side Length

• Depth Remarks Start Stop NO INDICATIONS FOUND
• Indication Length - As developed on outside surface of shroud. (Using 1.923*/Deg.)

                                                                               ~.
                                                                                                                                            .e
                                 . __ __ _.____ _ ___-_ _ _ ___ _                                    A--  _ _ - _ _ - _ _ _ _

PECO ENERGY COMPANY F PEACH BOTTOM ATOMIC POWER STATION , UNT 2 .

SUMMARY

OF CORE SHROUD EXAMINATION RESULTS I 2R10 i i i Weld No. H3 TABLE 1.3

                                                                                                                                                                                                         'I Aximuth Location Indication #  Surface    Weld Side                                                                   _ength*       Depth                                      Remarks Start                  Stop i
                                                                                                                      **                     Detected by 45'S/60*RL 1          ID        Lower                    8.34'                   9.33"                            1.79"                                                                                       :
                                                                                                                      **                     Detected by 45*S/60*RL 2          ID        Lower                   11.37*                  14.78*                           6.16"                                                                                        i
                                                                                                                      **                     Detected by 45'S/60*RL 3          ID        Lower                   41.04*                  41.70*                            1.19"                                                                                       !

4 ID Lower 70.32* 72.46* 3.87" ** Detected by 45'S/60*RL j

                                                                                                                      **                     Detected by 45'S/60*RL 5          ID        Lower                   73.63*                  75.24*                           2.91*                                                                              .

j

                                                                                                                      **                     Detected by 45'S/60*RL 6          ID        Lower                   79.75*                  81.20*                           2.62*
                                                                                                                      **                     Detected by 45'S/60*RL 7          ID        Lower                   82.40'                  83.61*                           2.19"                                                                                        l i

6.96* ** Detected by 45'S/60*RL 8 ID Lower 102.47* 106.32* l 118.30' 7.27" ** Detected by 45*S/60*RL 9 ID Lower 114.28* j

                                                                                                                      **                     Detected by 45'S/60*RL 10          ID        Lower                  120.79*                 123.25'                          4.45"
                                                                                                                      **                     Detected by 45'S/60*RL 11          ID        Lower                  142.76*                 144.41*                          2.98"                                                                                         {

3.14" ** Detected by 45*S/60*RL 12 ID Lower 148.96* 150.70*

                                                                                                                      **                     Detected by 45'S/60*RL 13          ID        Lower                  151.89*                 154.04*                          3.89"                                                                                         j i
                                                                                                                      **                     Detected by 45'S/60*RL 14          ID        Lower                  159.49*                 160.31*                           1.48"                                                                                         j t
                                                                                                                      **                     Detected by 45'S/60*RL 15          ID        Lower                  190.65*                 194.18*                          6.38"                                                                             ,
                                                                                                                      **                     Detected by 45'S/60*RL 16          ID        Lower                  239.18*                244.02*                          8.75"
                                                                                                                      **                     Detected by 45*S/60*RL 17          ID        Lower                 255.65*                 2$5.98*                          0.60*
                                                                                                                      **                     Detected by 45'S/60*RL 18          ID        Lower                 256.17*                 256.65*                          0.87*                                                                                          l I

19 ID Lower 344.11* 344.65* 0.98* Detected by 45'S/60*RL ' Indication Length - As developed on outside surface of shroud. (Jsing 1.80T/Deg.J ** Depth sizing not f,idvinsad f

PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION UNIT 2 - -

SUMMARY

OF CORE SHROUD EXAMINATION RESULTS 2R10 TABLE 1.4 Weld No. H4 Azimuth Location , Indication # Surface Weld Side Length

• Depth Remarks l Start Stop  ;

Upper 21.30* 21.91* ** Detected by creeping wave UT only. 1 OD 1.10* 66.96* 67.24' ** Detected by creeping wave UT only. 2 OD Lower O.51* l Upper 138.37' ** Detected by creeping wave UT only. 3 OD 138.81* 0.80* Upper 146.78* 149.97* 5.76" ** Detected by 45*S/60*RL 4 ID

                                                                                                                                              **                                                                           i 5                                 OD                                             Lower    151.88*     152.27*     0.70*                         Detected by creeping wave UT only.

153.32' 153.81* ** Detected by creeping wave UT only. 6 OD Lower 0.89*

                                                                                                                                              **          Detected by creeping wave UT cnly.

7 OD Upper 193.70* 194.03* 0.60* i Upper ** Detected by creeping wave UT only. 8 OD 197.22* 197.83* 1.10"

• Indication Length - As developed on outside surface of shroud. (Using 1.807*/Deg.) ** Depth sizing not performed i

i 5

                                                                                                                 ~,

PECO ENERGY COMPANY , PEACH BOTTOM ATOMIC POWER STATION UNIT 2 1

SUMMARY

OF CORE SHROUD EXAMINATION RESULTS 2R10 TABLE 1.5 , i Weld No. HS Azimuth Location , 1 Indication # Surface Weld Side Length

• Depth Remarks  :'

Start Stop 2.28' ** Detected by creeping wave UT only. 1 OD Lower 120.73* 121.99* I

• Indication Length - As developed on outside surface of shroud. (Using 1.80T/Deg.) ** Depth sizing not performed s

l

PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION UNIT 2 . .

SUMMARY

OF CORE SHROUD EXAMINATION RESULTS 2R10 TABLE 1.6 Weld No. H6 - Azimuth Location Indication # Surface Weld Side Length

• Depth Remarks Start Stop 1 ID Upper 10.90* 13.60* 4.73* .45' Detected by 45'S/60*RL
• Indication Length - As developed on outside surface of shroud. (Using 1.75"/Deg.)

i t t e i

                                                                                                                                                                                                       ?

PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION UNIT 2 s

SUMMARY

OF CORE SHROUD EXAMINATION RESULTS 2R10 t TABLE 1.7 Weld NO. H7 I Azimuth Location Indication # Surface Weld Side Length

• Depth Remarks i

NO INDICATIONS FOUND

• Indication Length - As developed on outside surface of shroud. (Using 1.754*/Deg.)

L i _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ _ . _ _ _ _ _ _ _ _ _ . _ _ - _ _ _ _ _ _ _ _ __ __ - ._ . _ _ _ u _ _ _ - e- -T

PECO ENERGY COMPANY

            ,                                             PEACH BOTTOM ATOMIC POWER STATION UNIT 2 REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT 2R10, October 1994 Docket No. 50 277 ATTACHMENT 4 l

t l l l l l l l l l l 1

 ,,       _-         - - - _ _ . - - - . - -                                            . _ - . -       . . -               . - . _ . . . . - . .        - ~ - . - .              . . . -

Structural Integrity Associates, Inc.

        ~

3315 AJmaden Expressway suite 24 san Jose,cA e51181557 October 14,1994 men : a.oru2co l RAM-94-333 w m s7sassa SIR-94-111, Rev. 0 l Mr. Vijay M. Nilekani l PECO Energy Company 965 Chesterbrook Blvd.,63C-9 Wayne, PA 19087-5691 l s

Subject:

Evaluation of Peach Bottom, Unit 2 Shroud Indications

Dear Vijay:

l StructuralIntegrity Associates (SI) has performed an evaluation of the flaw indications found , l during the inspection of circumferential welds H1, H2, H3, H4, H5, H6, and H7 at Peach l l Bottom, Unit 2, in order to determine the structural margin in each of these welds. The l l evaluation performed and documented herein was designed to evaluate operation without I i repair of these welds for an additional 24-month operating cycle for welds H1, H6, and H7, I and for two additional 24-month operating cycles for welds H2, H3, H4 and H5. The evaluations were performed following the approach used in the BWR Vessel and Internals l Project (VIP) evaluation guidelines [1] and the GE screening criteria (2] developed for l Peach Bottom, Unit 2, based on limit load and linear elastic fracture mechanics (LEFM) l techniques. The following sections of this letter report describe the methodology used to evaluate each weld, and the resulting safety margins. Inspection and Evaluation Methodology The inspection and evaluation approach employed at Peach Bottom, Unit 2 provides the j necessary information' for determin~ation of the allowable flaw lengths, including the l appropriate amount for crack growth and nondestructive examination detectibility l l uncertainties, for all observed flaws. Ultrasonic testing (UT) techniques were utilized which ( provided complete through-thickness determination of any observed indications. Due to l accessibility limitations, the extent of examination of these welds varied. In most cases, sufficient weld length was adequately interrogated to quantifiably demonstrate the condition, l and hence the structuralintegrity of the weld. However, in some cases, the extent of weld l length accessible for examination was limited. In these cases, the results of the examinations I l were extrapolated over the portions of the weld which were not examined, thus resulting in l a qualitative evaluation of the condition and structuralintegrity of the weld. Because both quantitative and qualitative assessments have been performed, evaluation ! methodologies will be addressed separately with respect to structural integrity evaluation. I In compliance with the evaluation guidelines [1], evaluation for two cycles of operation is l l l Anres. Dit $her Spring. BID FL Las6ertale, FL faipel, nnraa latemeines, lac, utsl}8 Psone 218-8644836 Phcrit.301 !89 2373 P C*e 3C5-J$41822 hone C2 333 !C8 Sher $ormg. MO Nortress. GA Pto e 331,589 2M0 Peore J0J7434'M I

      ~               -.                        . - - _ - - - - -               ,     -          , .         , . _       -                        -   . . . , .                 -

l l . Mr. Vijay M. Nilekani October 14,1994 Page 2 RAM-94-333/ SIR-94-111, Rev. O considered for those welds which can support a quantitative assessment (welds H2, H3, H4, and H5), and one cycle of operation is considered for those welds in which only a qualitative l assessment was performed (welds H1, H6, and H7). For purposes of this evaluation, one i cycle of operation is assumed to be two years in length, at 110% of current rated power and 110% of rated core flow. Acceptance Criteria The core shroud is a core support structure which provides lateral support for the fuel. The applicable codes, standards and classifications for the core shroud are as follows: The core shroud is classified as a safety-related component. The core shroud is not an ASME Code, Section III component. However, the original design is in accordance with the intent of Section III of the ASME Code. The evaluation of the core shroud was performed in accordance with the requirements of the BWR VIP's evaluation guidelines [1]. Flaw Evaluation Results Following completion of the inspection of the H1, H2, H3, H4, H5, H6, and H7 welds, flaw analyses were performed to demonstrate that the structural margins identified in the i evaluation guidelines were maintained for the actual flaw configurations which were l identified. (It should be noted that welds H2 and H7 have no reported indications.) The I flaw analyses were performed using limit load as the failure criterion for each of the welds. For the quantitative assessment, the evaluation performed here takes into account the distribution of uncracked material around the circumference of the shroud. In addition, the H3 and H4 welds, which are the most highly irradiated, were also evaluated using LEFM fracture methodology to be consistent with the evaluation guidelines [1]. For those welds, which could only be qualitatively assessed due to access limitations (welds H1 and H6), a different approach was utilized. The maximum flaw depth observed was assumed to be continuous for 360 of the shroud circumference. Because welds H1 and H6 are in areas 2 ofless radiation (< 3 x 10" n/cm ), only limit load techniques were utilized. In addition, an allowable flaw depth for weld H7 was determined using the same methoc: ology. Substantial conservatisnas were built into the flaw evaluation to account for the weld area examined, the weld area which was not examined, evaluation guidelines' detection and sizing ! uncertainties, through the thickness (radial) crack growth and circumferential crack growth, j and the evaluation guidelines flaw proximity criteria as applied to adjacent flaws. The speci6c conservatisms utilized in this evaluation are as follows: l f StructuralIntegrity Associates. Inc.

Mr. Vijay M. Nilekani October 14,1994 Page 3 RAM-94-333/ SIR-94-111, Rev. 0

1. A bounding crack growth rate (5x108 inches / hour) [1, 2] through-wall and around the circumference was applied to the cracks detected. For the circumferential crack growth, two cycles were utilized for the structural margin assessment.

2. For the quantitative assessment, allinspected regions which are identified as cracked are treated as through-wall cracks and assumed to grow by 1.6 inches l at each end during the next two operating cycles.

3. Per the evaluation guidelines, twice the shroud thickness, or 4 inches, was added to each end ofidentified indications.

l 4. For the quantitative assessment, all areas not inspected are assumed to be l cracked through-wall. These areas are also increased in length by the crack growth and uncertainty factors described above. l l S. For the qualitative assessment, identified flaws are increased in depth by 0.3 inches to account for UT sizing uncertainties, and 0.8 inches for one cycle of crack growth. This is in compliance with the evaluation guidelines. ) 6. For the qualitative assessment, the deepest identified flaw is assumed to be 360 in circumferential length. { l 7. ASME Code, Section XI proximity rules for adjacent flaws were applied.

8. ASME Code, Section XI pressure boundary safety margins were applied to l these evaluations even though the core shroud is not a primary pressure boundary. l The conservative assumptions described above were applied to each of the horizontal welds l examined in this report. Table 1 presents the membrane and bending stresses which were used for the limit load analyses for each of the welds identified in the table, and for the LEFM analyses performed for welds H3 and H4. One notes from Table 1 that the highest stresses are observed at the H7 weld, and the lowest stresses occur at the H1 weld location.

l The limit load analysis was performed for all welds evaluated in this study, the H1, H2, H3, H4, H5, H6, and H7 welds, and LEFM was also performed for the H3 and H4 welds. Because the Faulted Condition factor-of-safety is half the Upset Condition factor-of-safety, l it is clear that the loading condition which governs the analysis for welds above the core plate is the Faulted Condition, with the Upset Condition governing welds H6 and H7. i l Table 2 presents the results of the ultrasonic examination for each of the horizontal welds quantitatively evaluated in this report. For welds H1 and H6, which are qualitatively assessed due to examination accessibility limitations, the maximum detected flaw depth is increased by 0.3 inches for detection uncertainty, and 0.8 inches for one cycle of crack growth. The postulated resulting 360' circumferential flaws which were analyzed are 1.84 f StructuralIntegrityAssociates,Inc.

Mr. Vijay M. Nilekani October 14,1994 Page 4 RAM 94-333/ SIR-94-111, Rev. O inches deep for weld H1, and 1.55 inches deep for weld H6. Using the same methodology, it was determined that the H7 weld could satisfy acceptance criteria with an initial 0.8 inch deep flaw. The results of the limit load analysis for each of the horizontal welds examined is presented in Table 3, based upon the stresses reported in Table 1, and the examination results reported in Table 2. One observes from this table that the factors-of-safety for the Faulted Condition for welds H1, H2, H3, H4, and H5 range from 5.4 for weld H1 to 40.2 for weld H2. This compares to an ASME Code minimum factor-of safety of 1.4 specified for pressure boundary components under faulted loading conditions. For weld H6, the reported factor of safety for Upset Conditions is 14.0, with an allowable of 2.8. One should note that  ! the conservatisms utilized in this study are as described previously in this section. Finally, an evaluation of the H3 and H4 welds was performed using the LEFM methodology i to determine the applied stress intensity factor resulting from the conservatively estimated cracking combined with the bounding loading condition (the Faulted Condition) for these weld locations. The results of this analysis demonstrate that the 150 ksi-(in)M toughness which is presented in the evaluation guidelines (1) as the acceptable fracture toughness for this material under irradiation embrittled conditions is met. Table 4 illustrates that the ASME Code minimum factor-of safety of 1.4 has been met under this loading condition for the flaws present at welds H3 and H4. Summary Based upon a review of the examination data for circumferential welds H1, H2, H3,li4, H5, H6, and H7, there is substantial margin for each of these welds under conservative, bounding conditions to allow for continued operation for a minimum of one additional 24-month operating cycle for welds H1, H6 and H7, and two additional cycles for welds H2, H3, H4 and H5. The analyses performed included limit load analyses under bounding design basis conditions, and LEFM for the postulated highest fluence welds. The quantitative evaluations were performed with the assumption that all regions uninspected were cracked through wall, and that any cracking observed was cracked through wall. Additionally, all areas assumed to be cracked were grown (at each end) at the bounding crack growth rate of 5x105 inches / hour for two cycles, and increased in length by 4 inches (for a total increase of 5.6 inches at each end). The qualitative evaluations were performed with a 0.3 inch sizing uncertainty and 0.8 inches for one cycle of crack growth. ASME Code safety margins were used, and were significantly exceeded in all cases for the specified operating period. Very truly yours, R. A. Mattson, P.E. Arsociate attachments ( StructuralIntegrity Associates, Inc.

 ,      7 -. _ - . . _ _

l l ' Mr. Vijay M. Nilekani October 14,1994 l Page 5 RAM-94-333/ SIR 94-111, Rev. O t l References 1 l

1. GE Nuclear Energy,"BWR Core Shroud Inspection and Flaw Evaluation Guidelines",

GENE-523-113-0894, September 1994. l l 2. GE Nuclear Energy," Evaluation and Screening Criteria for the Peach Bottom Unit 2 Shroud", Report Number GENE-523-1761293, December 13,1993. ) l 1 e i e d

                                                              =

f StructuralIntegrity Associates. Inc.

Table 1 - Membrane / Bending Stresses Shroud Stresses 2 Weld Location Membrane Bending Upset Faulted Upset Faulted H1 0.38 ksi 0.88 ksi 0.09 ksi 0.18 ksi H2 0.38 ksi 0.88 ksi 0.12 ksi 0.23 ksi H3 0.36 ksi 0.83 ksi 0.14 ksi 0.27 ksi H4 0.36 ksi 0.83 ksi 0.27 ksi 0.55 ksi H5 0.36 ksi 0.83 ksi 0.42 ksi 0.84 ksi H6 0.62 ksi 1.10 ksi 0.46 ksi 0.91 ksi j H7 0.62 ksi 1.10 ksi 0.56 ksi 1.11 ksi l

                                                                                                                                                                /

NOTE: 1. Because the Faulted Condition factor of safety is half the Upset Condition factor-of-safety, it is clear that the Faulted Condition  ; governs for welds H1 through H5, and the Upset Condition i governs for welds H6 and H7. l ) Attachment to l

'            RAM-94-333/ SIR-94-111                                                                                                                               .

Structural Integrity Associates, Inc. i

                   ,n                                  - - . . - . - -                  -.                                -                    -              -     -        -  - - - - - - . - . - _

Table 2 I UT Examination Results Used for i Quantitative Analyses l 1 - Weld Unflawed Material for Analyses Total Unflawed Location Material H2 11.37*-154.31 , 191.37*-333.31 537"/79 % i H3 17.94 -37.88 , 44.86 -67.16*, 342"/54 % i 86.77 -99.31 , 126.41*-139.60 , 197.34 236.02 , 247.18 -252.49 , 259.81 -340.95

H4 12.15 -18.14 , 25.07* 63.80 , 449"/70 %

l 70.40*-135.21 , 156.97 160.53 , l l 200.99 -341.03* l ! HS 12.15 -117.57 , 125.15 -153.53 , 4 J4"/74% l 199.65 -333.53 NOTES: 1. All reported indications have been assumed through wall and increased in length by 5.6 inches (2T + 2CG) at each end.

                                                                                                                  /

2. All areas not inspected are assumed to be cracked through wall, and extended in length by 5.6 inches (2T + 2CG) at each end. l l

l l l l i i Attachment to RAM-94-333/ SIR-94-111 . StructuralIntegrity Associates, Inc.

_. . _ - - _ _ . - _ . _ . ~ - - _ . _. _ . ._. - . _ . . . __ l l Table 3 l Limit Load Factors-of Safety l l Weld Location Factors-of-Safety22 l ! HI 5.4 H2 40.2 H3 26.3 H4 27.3 HS 24.2 H6 14.0 H7 3.12 NOTES: 1. The Faulted Condition governs for welds H1, H2, H3, H4, and H5, with an allowable factor of safety of 1.4.

2. The Upset Condition governs for welds H6 and H7, with an allowable ,

factor of-safety of 2.8.

3. Based upon an assumed initial flaw depth of 0.8 inches.

l i Table 4 LEFM Factors-of-Safety Weld Location Factors-of-Safetyi H3 3.6 H4 2.8 I NOTE: 1. The Faulted Condition governs, with an allowable factor-of safety of l 1.4. 1 Attachment to RAM-94-333/ SIR-94-111 StructuralIntegrity Associates. Inc. 1

St;ti:n Suywrt C urtm_nt GL 94-03

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_7

• PECO ENERGY F..R, !a ..L ,

                                                                            %$ cms't*:tco" & u.x 
                                                                            ?la<ne FA 'OSNN' November 3,1995 Docket No. 50 278 Ucense No. DPR 56 U. S. Nuclear Regulatory Commission Attn: Document Control Desk                                                                       l Washington, DC 20555

Subject:

Peach Bottom Atomic Power Station, Unit 3 ' Supplemental Response to Generic Letter 9443 l Summary of Core Shroud Inspection Results l l Dear Sir . 1 In our letters from G. A. Hunger, Jr. (PECO Energy Company) to U. S. Nuclear Regulatory Commission (USNRC), dated August 24,1994 and June 16,1995, PECO Energy Company , provided inspection plans for the Peach Bottom Atomic Power Station (PBAPS), Unit 3 core l shroud. These plans were submitted in accordance with Reporting Requirements 1 and 2 of l Generic Letter (GL) 94 03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors." By letter dated October 25,1995, the USNRC Indicated that the proposed l scope of inspections was acceptable. The purpose of this letter is to provide the final summary report, as requested by Reporting Requirement 3, of the GL in summary, the overall results of the inspection revealed a moderate amount of indications. Less than 12% of the examined weld length was found to contain flaws. The evaluation of the results was performed following the approach outlined in the "BWR Core Shroud inspection and Flaw Evaluation Guidelines," GENE 523115-8094, Revision 1, dated March 1995. This evaluation, based on the examination data, concludes that there is a substantial margin for each of these welds under conservative, bounding conditions to allow for continued operation of PBAPS, Unit 3. If you have any questions, please contact us. Very truly yours. L[}(. d k 1# G. A. Hunger, Jr., Director Licensing Attachment, Affidavit cc: T. T. Martin. Administrator, Region I. USNRC W. L Schmidt, USNRC Senior Resident inspector, PBAPS V$ ) '

1

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COMMONWEALTH OF PENNSYLVANIA  : I

ss.

COUNTY OF CHESTER  : 1 D. B. Fetters, being first duly sworn, deposes and says' l l That he is Vice President of PECO Energy Company; that he has read the enclosed supplemental response to Generic Letter 94-03, for Peach Bottom Facility Operating License DPR 56 and i knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief. l 1 m \ N' - m h o X\ .Lw J l Vice President I l l l Subscribed and sworn to before me this & day l 1 of %ggyM 1995. l l l 0AlA L. ' U O Notary Public scrac seal iVayne H "Act Notary Ptbic 1pf '.* O'esterComty MyConwne+* :39 :5* May13.1996 Meneer,PerC; Cih.caton of Notanes

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a e O l l l l l l 1 ATTACHMENT 1 1 1 1 1 1 1 I l 1 l l l 1 l l l 1 I l l l I i l

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PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION

 '                                                     UNIT 3 REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT 3R10, October 1995 Docket No. 50 278 in September and October of 1995, during the tenth refueling outage of Peach Bottom Atomic Power Station (PBAPS), Unit 3, the core shroud structure was comprehensively inspected. These inspections were conducted to determine the condition of the shroud welds, relative to the potential for existence of Intergranular Stress Corrosion Cracking (IGSCC). The effort satisfied the commitments made for PBAPS, Unit 3,in the PECO Energy response to NRC Generic Letter 94-03, dated August 24,1994, and as discussed in our PBAPS, Unit 3 core shroud inspection plan, forwarded to the NRC in our letter dated June 16,1995. The inspections were conducted in accordance with the guidance provided by the Bolling Water Reactor Vessel and Internals Project             '

(BWRVIP), as presented in the "BWR Core Shroud inspection and Flaw Evaluation Guidelines", l GENE 523113-0894, Rev.1, dated March 1995 (Reference 1). The following describes the overallinspection effort and summarizes the results of this effort. BACKGROUND: The PBAPS, Unit 3 shroud was fabricated by Rotterdam Drydock Co. LTD., Rotterdam, Holland. l The product forms used for this fabrication included 2" thick ASTM A240, Type 304 stainless steel plate (for shroud cylinders), and ASTM A182, Grade F304 seamless, stainless steel rolled forgings (rings). The plate materials contain relatively high carbon contents (.059% to .062%), while the ring forgings contain lower carbon contents (.030% to .035%). The product forms where joined using the submerged are welding process. The weld filter metal used was ASTM A371 Type Er308, with low carbon content. Welds H 1 through H 6 were welded from both surf aces, using a double bevel  : weld prep. Weld H 7 was welded from the inside surface of the shroud using a single bevel weld l prep and a backing ring. The H 7 weld was made at the PBAPS site, and it attached the pref abricated shroud structure to the Reactor Pressure Vessel. This weld is a dissimilar metal weld (304 stainless to Alloy 600). The filler metal used for this weld was ASTM B 304 Type ERNiCr 3 (Alloy 82). The process used for this joint was the Shielded Metal Arc Welding process. Attachment 1 includes a drawing which depicts the shroud configuration, weld locations, and materials of fabrication. The PBAPS, Unit 3 shroud has been in service since December 1974. During the first decade of hot operation, PBAPS, Unit 3 operated with relatively high primary water conductivity. Unit 3's arithmetic mean conductivity exceeded 1.0 S/cm during the first few years of operation. Subsequently, conductivity values were steadily decreased to below current EPRI guidelines.1992 and 1993 values were actually less than 0.1 pS/cm. The effects of such early water chemistry history on the susceptibility of the shroud welds to IGSCC are addressed in Reference 1. The above described factors place the PBAPS, Unit 3 shroud into inspection Category C, as defined by Reference 1. This category has a high potential for some amount of shroud cracking. and, therefore, comprehensive inspections of welds H 1 through H 7 are recommended. Page 1 of 4

PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION UNIT 3 REACTOR PRESSURE VESSEL CORE SHROUD INSPECHONS FINAL REPORT 3R10, October 1995 Docket No. 50-278 INSPECTIONS: The scope of the core shroud inspections included all of the shroud circumferential welds (e.g. H 1 through H-7). The method used for inspection of these circumf erential welds was Ultrasonic Testing (UT), performed from the outside surf ace of the shroud, using the General Electric Nuclear Energy (GENE) SMART 2000 data acquisition system and the GENE OD Tracker. This shroud Inspection equipment was satisfactorily demonstrated at the EPRI NDE Center. The extent of the planned inspections included all portions of the circumferential welds which were accessible for the above described equipment. This scope and extent of planned inspections was identified in PECO Energy's second response to Generic Letter 94-03, dated June 16,1995. 1 The UT scanning was accomplished using three transducers. These transducers included 45' shear wave,60' longitudinal wave, and creeping wave units. The transducers scanned each Heat Affected Zone (HAZ) of the accessible lengths of each weld. The creeping waveiransducer was used to enable better near surf ace detection capabilities. i 1 The purpose of the shroud inspections was to assess the condition of the shroud circumferential 4 l welds so that the integrity of the shroud structure could be quantitatively demonstrated. Additionally, j ' the inspection results will be used to establish a baseline of this condition for comparison to future  ! inspection results. This baseline data and subsequent inspection results will also be used to develop schedules for future shroud inspections, evaluations, or repairs. The extent of shroud weld inspections performed during 3R10 include: 84.5% of the length of Weld H 1, 584" 84.5% of the length of Weld H-2. 584" 89.5% of the length of Weld H 3, 582" 1 89.2% of the length of Weld H-4, 580" l 90.8% of the length of Weld H 5, 591" j 80.1% of the length of Weld H-6, 506" 89.6% of the length of Weld H 7. 4 566" ' Subtotal 3993" x 2 (HAZ per weld) , Total 7986" The extent of these weld inspections is graphically depicted on the attached weld maps for welds H 1 through H 7, (Attachment 2). l i i Page 2 of 4

PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION

 ,                                                           UNIT 3 REACTOR PRESSURE VESSEL CORE SHROUD INSPECVONS FINAL REPORT                                                   l 3R10, October 1995                                               l Docket No. 50 278                                               '

RESULTSt l A sufficient length of each circumferential weld was inspected to quantifiably demonstrate the condition and, therefore, the structural integrity of these welds. i Some indications were found on welds H 1, H 3, H-4, and H 5. No indications were found on welds H 2, H 6, and H 7. The general beation of the indications are depicted on the attached weld maps (Attachment 2). Shroud Weld Indication Data Sheets provide details of the as-found indications, and are included as Appendix 1 of Attachment 3. EVALUATIONS: All as found indications were assumed to be through wall. Therefore, depth sizing of the indications was not utilized. Additionally, the weld lengths which were not inspected, due to inaccessibility. were also assumed to be through wallindications. Inspection results were initially compared against a screening criteria, which had been developed prior to the inspections. Application of this very conservative screening critena allowed for a rapid assessment of the acceptability of each weld, based on initial examination data. The screening was applied for both the Limit Load and Linear Elastic Fracture Mechanics Methodology. If the J results of this screening indicated that sufficient unflawed material existed, the weld was considered  ! acceptable. Ultimately, a detailed evaluation was performed for all welds,to determine the margin i of safety for each weld (see Tables 2 3 through 2-6 in Attachment 3). The detailed evaluations were performed by General Electric Nuclear Energy. These evaluations used the guidance provided in the evaluation portion of Reference 1. The as-found Indication lengths were adjusted for upper bound crack growth. NDE uncertainty (0.4" plus 0.5' each end), and proximity factors. The resultant Indication lengths (as evaluated indications) were then used to calculate the amount of safety margin remaining in the subject weld, using the limit load methodology. Additionally, for Welds H-3 and H-4, the Linear Elastic Fracture Mechanics (LEFM) technique was used, due to the extent of neutron exposure received at these weld locations. The safety factors were calculated against the most limiting design basis loading conditions, derived from the General Electric Nuclear Energy Screening Criteria Document (Reference 2) and the PBAPS, Unit 3 UFSAR. The loadings also considered Power Rerate conditions and updated seismic loadings. A more detailed discussion of the evaluations, including factors utikzed for crack growth and NDE uncertainties,is contained in the GENE Evaluation Report GENE-523-A104 0995, (Attachment 3). CONCLUSIONS: A 10CFR50.59 determination and safety evaluation has been developed and reviewed by the Plant Operations Review Committee (PORC). The conclusion of this evaluation indicates that no unreviewed safety questions exist as a result of the shroud inspection findings. Page 3 of 4

PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION

• UNIT 3 REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT i 3R10, October 1995 Docket No. 50 278 The results of the inspections and evaluations conclude that the condition of the PBAPS, Unit 3 shroud, projected through the next two operating cycles, will support the required safety margins, specified in the ASME Code and reinforced by the BWRVIP recommendations. Additionally, the results of these UT inspections substantiate the use-as-is disposition of NCR No. 93-00743, Rev.

1, developed during the PBAPS, Unit 3 Refueling Outage 9 (1993), as a result of shroud visual inspections findings, and the Safety Analysis developed in response to Generic Letter 94-03. The extent of the shroud inspections provide a comprehensive baseline for comparison to future inspections. PECO Energy will continue to follow the developments of the BWRVIP guidance documents, and will evaluate their applicability to the PBAPS Site. Reinspection of the shroud welds will be determined following resolution of the BWRVIP reinspection recommendations.

REFERENCES:

1. BWR Core Shroud Inspection and Flaw Evaluation Guidelines, GENE-523113 0894, Rev.

1, March,1995. l

2. Screening Criteria and Flaw Evaluation Methodology for the Peach Bottom Unit 3 Shroud. I GENE 523 A076-0895, September,1995.

l

3. Evaluation of the Peach Bottom Unit-3 Core Shroud Indications (Refuel 10), GENE 523-A104-0995, Revision 1, October 1995. l

4. BWR VIP Core Shroud NDE Uncertainty & Procedure Standard, dated November 21, i 1994.

5. NRC Safety Evaluation of Referenced Documents 1 and 4, dated June 16,1995.

Page 4 of 4

PECO ENERGY COMPANY PEACH COTTOM ATOMIC POWER STATION UNU 3 REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT l 3R10, October 1995  ; Docket No. 50-278 ATTACHMENT 1 , 1 REACTOR PRESSURE VESSEL - SHROUD PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 70RYER/ SEPARATOR SUPPORT RING WELO NO. \ 1

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i 1 i PECO ENERGY COMPANY ! PEACH BOTTOM ATOMIC POWER STATION 1 UNIT 3 I ' REACTOR PRESSURE VESSEL CORE SHROUD INSPECTIONS FINAL REPORT 3R10, October 1995 I Docket No. 50 278 j ATTACHMENT 3 i 4 1 i  !

                                                                                       )1 1

i

GENE-523-A104 0995 Revision 1 DRF 137-0010-8 Evaluation of the Peach Bottom Unit-3 Core Shroud Indications (Refuel Outage 10) a October 1995 Prepared by: h 4 Nfarcos L. Herrera, Principal Engineer Engineering & Licensing Consulting Services

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K4lrv[< - - m Karina Flynshtein, Engineer Engineering & Licensing Consulting Services m _ GE Nuclear Energy San Jose, CA W e.s

GCSuteu Enwy GENE.5%U04.ons

                                                                      .                          He,ston i I

l IMPORTANTNOTICE REGARDING l CONTENTS OF THIS REPORT Please Read Carefully l 4 The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between PECO Energy Company and GE, and nothing containedin this document shall be construedas changing the contract. The use of this F- information by anyone other than PECO, orfor anypurpose other than thatfor which it is intended under such contract is not authori:ed; and with respect to any unauthori:ed use. GE makes a representation or warramy, and assumes no liability as to the completeness, accuracy, or u.sep ' ness of the information comained in this document, or that its use may not infringe prinnely owned rights. 1 I l* l 1. l 0 i

     ,   , GCNuclext Energy GENE.523.A104 0895
                                                             ~

Revision i Table of Contents EXECUTIVE

SUMMARY

1. INTRODUCTION 1 1.1 Flaw Disposition Approach 1 1.2 References 8

2. EVALUATION OF UT RESULTS 9 ,

2.1 References 16  ! l 1

3.

SUMMARY

AND CONCLUSIONS 17 l ~ l l i APPENDIX A UT EXAMINATION RECORDS i l

                                                                 .                           l m

i as

i , GE Nuclear Energy _

• GE.VE.323.A104 0393 Revision 1 EXECUTIVE

SUMMARY

UT inspection of the H1 through H7 core shroud welds was performed during refuel outage 10 at Peach Bottom Unit-3. Indications were observed in the inspected areas of welds H1, H3, H4, and H5. Indications were not observed at welds H2, H6, and H7. This report presents the results of the application of the screening criteria and flaw evaluation calculations for the observed UT detected indications. Structural margin is assured if the observed indications meet the screening criteria or if the calculated safety factors, using the flaw evaluation method, exceed the required safety factors. Screening criteria and flaw evaluation methodology were prepared in a previous analysis. The flaw evaluation needs to be performed if the flawed condition exceeds the screening criteria. Even if the screening criteria is met, based on assuming that all UT detected flaws are through-wall, it is appropriate to reevaluate the indications using the flaw evaluation methodology to demonstrate the actual structural margin. However, reconciliati6n using the flaw evaluation methodology is not mandatory to determine the actual structural margin or to justify continued operation. Both the screening criteria and flaw evaluation methodology use linear elastic fracture mechanics (LEFM) and limit load concepts to determine the acceptability of the flaws. The limiting flaw length, based on either LEFM or limit load, was used for the allowable flaw size at the H3 and H4 welds. This evaluation used a NDE uncertainty of 0.4 inches plus half a degree which was added to each flaw end. The results of this evaluation indicate that the screening criteria is satisfied at all weld locations. In addition, the flaw evaluation indicates safety factors well in excess of the required safety factors. Thus, structuralintegrity over the next two year  ! operating cycle is demonstrated. 1

 .                                                                                                          l l
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l t

    .    . GE Nuclear Energy GENE.323-A104.vt93 Redsion 1

1. INTRODUCTION l

This report presents the evaluation of the 1995 outage (Outage 10) ultrasonic test l inspection (UT) results for the Peach Bottom Unit-3 core shroud. Reference 1-1 I presented the core shroud screening criteria and flaw evaluation methodology for Peach Bottom Unit-3. The UT detected indications (See report sheets in Appendix A) were evaluated per the rnethodology and procedures presented in Reference 1-1. i The evaluation presented in this report (Section 1.1.1) uses the initial screening criteria methodology for circumferential welds along with LOCA and updated loads for seismic events. In addition, the flaw evaluation calculation (Section 1.1.2)is presented which can be used if the screening criteria is exceeded or if a closer estimate of the safety margin is desired. Section 1.1 describes the approach to disposition the indications using the two methods. 1.1 Flaw Disposition Approach , The approach in dispositioning the flaws in the Peach Bottom Unit 3 core shroud is outlined in this section. This approach is consistent with the approach taken to disposition indications at several other BWR plants since core shroud cracking has been observed and is consistent with the BWR VIP methods in Reference 1-2. Figure 1-2 shows a flow chart summarizing the process of shroud cracking disposition. The initial evaluation, based on the conservative screening criteria, is first performed. This conservative evaluation can be used to quickly disposition the indications based on many simplifying assumptions which clearly illustrate the conservative nature of this screening criteria. Two of these significant assumptions, which have been verified as such since 1993, are i) all indications are through-wall even though all detected indications were found to be part through-wall, and ii) all indications after application of the proximity rules are combined into one single indication which is oriented along the axis ofminimum

 ,          moment ofinertia.

A flaw evaluation may be performed if the as-found indications exceed the screening l criteria. This flaw evaluation can take into account the actuallocation and flaw 1 characterization from the UT inspection. Even if the indication meets the screening

i 1

e

1 l

     ,    , GENuclear Energy GENE.533.A104 0293 Rnnion !

criteria, it is considered prudent to determine the actual structural safety factor for the i, flawed condition. This information can also provide additional guidance for future planning and management of core shroud cracking. i ' The UT detected flaw lengths used in the screening criteria and flaw evaluation calculations included an uncertainty factor on length sizing. This uncertainty factor includes consideration for NDE technique uncertainty and NDE delivery system uncertainty, NDE length uncertainty values of 0.4 inches for NDE method plus half a degree for the delivery system (Reference 1-3) were added to each flaw end in this evaluation. This is a very conservative approach, considering the basis and the latest uncertainty data available from the BWR-VIP (Reference 1-4). The delivery system uncenainty value of half a degree applies only to longer indications which require transversing of the tracker delivery device to locate each end of the indication. The uncertainty value for short flaws (not requiring tracker movement) is actually very small. The larger uneenainty value was applied to all identified indications, regardless of identified length. The latest BWR-VIP data for NDE technique uncenainties, which were derived from demonstrations at the EPRI NDE Center, reflect substantially lower values for the techniques utilized during the Peach Bottom Unit 3 examinations. Demonstrations #5 and

            #16 (Reference 1-4) indicate a NDE technique uncertainty value of zero inches.

j Nevertheless, the larger NDE uncertainty value was applied to maintain the maximum level of conservatism and to utilize data officially submitted to the NRC. l l l There are areas which could not be inspected during the UT inspection due to obstruction , by other components. In the calculations presented in this report, all uninspected areas j were assumed to contain through-wall flaws along the entire length of the uninspected l zone. The estimated crack growth and uncertainty were added to the assumed through- I wall flaws in the uninspected zones. This is likely a conservative assumption based on the UT results for all welds. Allindications were found to be pan-through-wall. s 1.1.1 Screening Criteria { , The guiding parameter used for the selection of the indications for further evaluation is the allowable through-wall flaw size, which aiready includes the stnactural safety factors. If all of the UT detected indications are assumed to be through-wall, then the longest flaws, or \- . 1 2

GENuclear Energy

      .    .                                                                                    GENE-523-A104-0395 ansion s combination of flaws, would have the limiting margin against the allowable through-wall flaw size. In reality, none of the indications are through-wall, and therefore, the criteria -

and methods presented for this method are conservative. The through-wall characterization of the indications can be incorporated in the flaw evaluation methodology which is described in Section 1.1.2. The result of this procedure will be the determination of the effective (limit load) and ! equivalent (LEFM) flaw lengths which will be used to compare against the allowable flaw sizes and selection ofindications for more detailed evaluation if necessary. The , determination of effective flaw lengths is based on ASME Code, Section XI, Subarticle IWA-3300 (1986 Edition) proximity criteria. These criteria provide the basis for the combination of neighboring indications depending on various geometric dimensions. The effective flaw lengths are summed into one single indication. This single indication is compared with the screening criteria allowable flaw size. Crack growth over a subsequent l two year operating and power rerate cycle is factored into the criteria. l The selection ofindications for further investigation can be performed by evaluating :he resulting effective flaw lengths. Indications with effective flaw lengths greater than the l allowable flaw sizes would require more detailed analysis such as the flaw evaluation method. The screening criteria procedure described here is conservative since all of the indications are assumed to be through-wall and are being compared against the allowable through-wall flaw size. A summary of conservatisms used in the screening criteria analysis is presented in Table 1-1.  ! l { l 1 r l l o , l 1. 1 I l 3

     . GENuclear Energy                                                          GE.VE.533.A104 0893 Revision i
       ,                Table 1-1 Conservatisms Included In Screening Evaluation

1. All surface indications were assumed to be through. wall for this analysis

2. Allindications are assumed to be grouped together for the limit load calculation and no credit is taken for the spacing between indications.

3. ASME Code primary pressure boundary safety margins were applied even though the shroud is not a primary pressure boundary.

4. ASME Code, Section XI proximity rules were applied.

5. An additional proximity rule which accounts for fracture mechanics interaction between adjacent flaws was used.

6. Both LEFM and limit load analysis were applied, even though LEFM underestimates allowable flaw size for austenitic materials and is not required per ASME Code Section XI procedures.

7. Fracture toughness measured for similar materials having a higher fluence was used.

8. The bounding crack growth estimated for the subsequent fuel cycles was included in flaw lengths used for evaluation. -

9. A bounding NDE uncertainty factor was included in the flaw lengths used for evaluation.

~                                                                                                        l
                                                                                                         \

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         .            < GE Nuclear Energy GENE.383.AlN.n93 Reklon 1                      '
             ,           1.1.2 Flaw Evaluation The flaw evaluation method can take into account the indication characterization information provided by the UT inspection. Specifically, the azimuthallocation and depth of the indications can be taken into account when determining the structural safety factor.

Crack growth over an operatirig cycle of two years and power rerate is factored into these calculations. For purposes of this evaluation, all detected flaws and uninspected areas were assumed to be through-wall flaws. l The flaw evaluation methodology (Reference 1-2) can include the assumption of through-

l. wall or part th ough-wall indications. Both limit load and LEFM are considered in this evaluation. For limit load, analysis can be performed for a random distribution of -

  ,                       indications varying in length and depth. In addition, uncracked ligament can also be modeled. The limit load allowable flaw length is defined for the given applied loads. The net-section stress equals the flow stress of the material at the flawed section (with applicable safety factor).

The LEFM evaluation considers the interaction of neighboring indications to establish an equivalent flaw length. The LEFM allowable flaw length is defined when the applied

   ,                      stress intensity factor equals that of the material fracture toughness.

l

                                                                                                                                                                            ]

l l" l J l 5

                                     - - - . va-,. , - . . . -

e .- -- - - - ~r: -,. . - - _ - - - ,-,,-........--,-..----4,.. . - - n,-. - - . -

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           .              .                   GENuclear Energy GENE 523.A104 0895
              =                                                                                                         .

Rethion J I l j Shroud Head Suppon Ring l H1 ' H2 - TCPGuide Suppen Ring NH3

                                                                                                        'H4 H5
                                                                                                        /                       Core Plate Suppon Ring y
                                                                                                             \        H6 Shroud Suppon Plate _

H7 v u.i H8 _ t Figure 1-1 Schematic of Core Shroud Welds l ' I i b 6 e

  - - -                                         e ------    , , _.. , . - - - , , , ,       ,     m         , _ . , .           ,--,--..,m .    - - , - . . - - - . , .       . .  -,,-~<y,      .m-~ m-. - - - -

GENuclear Energy GENE.323.A104 0895 Rohion ! UT Results ' Perform Screening Criteria l Is screening Yes Continued l Criteria Met?  : Operntion  ! Justified { l No I t v . . Perform Flaw Evaluation 1 1 l Safety Factors Yes Met?

                                                                                          ,r No                              . Perform Flaw Evaluation to Demonstrate Stmetural Margin Repair Required Figure 1-2 Flaw Disposition Procedure e

7 4 e

l GE Nuclear Energy GENE.323.A!040393

                                                            .                          Resblon i 1.2     References 1

1-1 Screening Criteria and Flaw Evaluation Methodology for the Peach Bottom Unit-3 Shroud Indications, GENE 523-A076 0895, DRF 137-0010-8, August 1995. 1-2 BWR Core Shroud Inspection and Flaw Evaluation Guidelines, GENE-l 13-0894, l DRF 137-0010-07, Rev.1, March 1995, Prepared for the BWR Vessel and Internals Project Assessment Subcommittee. l 1-3 BWR-VIP Core Shroud NDE Uncertainty & Procedure Standard, November 1994. l l-4 Reactor Pressure Vessel and Internals Examination Guidelines, BWR VIP (Draft) l Proprietary Report, September 1995. 6 8

 .      - . - -          _ - - - . _ _ . _ ,        -- --                  . . _   . ~ _ . . -           _ - = . _ . - - - _ _ . .                   . . . -

GENudear Energy. GENE.523.A1040393

     ~,                                                                              .

Reshfun 1

2. EVALUATION OF UT RESULTS This section provides the results of the application of the screening criteria and flaw evaluation methodology for the Peach Bottom Unit-3 core shroud circumferential welds.

 -                                                                                                                                                                1 The evaluation was performed.using a conservative approach. All uninspected areas were treated as through-wall' flaws. Crack growth for one c, cle and NDE technique and                                                               :

1 delivery system uncertainty were added to the end of eat indicatior In addition, all l indications were treated as being through wall. UT inspect 3n results indicate that all I indications are part through-wall. Appendix A contains the UT examination reports for welds H: through H7 Indications 1 l were not detected at welds H2, H6, and H7. Thus, welds H2, H6, and H7 were assumed ' l- to have through wallindications only in the uninspected regions. l l l, All indication lengths, including the uninspected area lengths, were increased by the l l assumed length uncertainty (0.4 inches plus a half a degree on length at each flaw end) plus two times the annual rate of crack growth for one 24 month operating cycle at each !~ flaw end. l- The stresses used for the flaw evaluation are shown in Table 2-1 (from Reference 1-1). Safety factors were calculated using the Distributed Ligament Length (DLL) computer

l. program (Reference 2-1). The procedure for evaluating the flaws for the screening criteria l

was: l

1 I

l-

1) Add crack growth for one cycle and length uncertainty to each flaw end for all flaws and uninspected area lengths from the UT examination repons (Appendix A).

2) Determine if flaws need to be combined based on proximity rules.

3) Sum all effective lengths.

4) Compare length sum to allowable effective length for limit load.

5) Determine equivalent length for any pair ofindications and compare to LEFM criteria.

Some of the observed indications at welds H1, H3, H4, and H5 were combined for this evaluation due to the added crack gromh and NDE uncertainty and due to the proximity l' criteria. Table 2-2 shows which indications were combined. L l > = 9 l-

GENuclear Energy

    '                                                                                      GENE.323.A104 0393
       ,                                                        .                                        Revklun 1 For the flaw evaluation calculations, the first two steps are identical to those for the                        ;

screening criteria. These flaw lengths (after proximity criteria application) are input into the DLL computer program which accounts for the azimuthal location of the indications (assumed to be through-wall).

                                                                                                                         \

i The calculated safety factors for both normal / upset and emergency / faulted conditions are l 1 shown in Table 2-3. It can be seen from Table 2-3 that there is a large safety margin 1 between the calculated and the required safety factors. Table 2-4 presents the calculated l total flaw lengths for the screening criteria. l 1 Weld H4 was found to contain an indication which is greater than 50% of the wall , thickness. Through-wall propagation of this indication cannot be ruled out. For an assumed fully circumferential flaw, Reference 2-2 indicates that the flow would occur , through a gap ofless than 0.002 inches. The estimated flow through such a gap would typically be about 0.05% of total core flow (based on a 0.002 inch gap around the shroud entire circumference and a typical pressure of eight pounds per square inch). Flow of this magnitude will have no impact on plant operation and will not be detectable. The observed indication at Peach Bottom Unit 3 at weld H4 which was found to be greater than 50% of the wall thickness is projected to grow to a length of 32 inches after one cycle of operation. This indication would then be 5% of the shroud entire circumference. Peach Bottom Unit 3 operates at a maximum pressure of 14.12 psi (Reference 2-3) during normal operation. Therefore, the expected leakage from a 1

~

through-wall flaw of this length would be less than 0.005% of the total core flow (this takes into account the higher operating pressure than the Reference 2-2 assumption). Therefore, the leakage through this indication would not be significant. 1 M 6. m 6 O

 .. -                                                     10

GENuclear Energy GENE.523.AIO4.cs95

         .e                                                          -

Reklon ]

          ,         Table 2-1. Primary Membrane and Bending Stresses at the Shroud Welds W eld                Normal / Upset         Emergency / Faulted Designation         P,,, (ksi)     P. (ksi)    P,, (ksi)       P (ksi)

HI 0.381 0.117 0.837 0.217 H2 0.381 0.I59 0.837 0.293 H3 0.359 0.186 0.787 0.340 H4 0.359 ' O.355 0.787 0.611 H5 0.359 0.535 0.787 0.944 H6 0.624 0.570 1.053 1.005 l H7 0.624 0.728 1.053 1.329 i Table 2-2. Combined Indications HI Indication #1 and Uninspected area from 340* to 11.20' Indications #5, #6, and #7 , H3 Indications #3 and #4 Indication #5 and Uninspected area from 169.75* to 189.20* Indications #8 and #9 Indication #10 and Uninspected area from 352.97* to 11.20' and Indication #1 1 H4 Indications #2, #3, #4, and #5 Indications #7, #8, #9, #10, #11, #12, and #13

    .                    Indications #19 and #20 Indications #21 and #22                                  .

Indications #23 and #24 Indications #27, #28, #29, #30, #31, and #32 Indications #34, #35, and #36

    ~

Indication #1 and Uninspected area from 349.82* to 9.40' HS Indications #2 and #3 Uninspected area from 351.20' to 9.20' and Indications #4. #5, #6, #7, #S. and #9 45 4 J

      .                                                          I1

t

     .    . GE Nuclear Enstgy                                                          GE.VE.5H.AlN.Ons
        .                                                         .                                Ression 1 Table 2-3. Flaw Evaluation Calculated Safety Factors (Required SF: 2.77 for Normal and Upset,1.39 for Emergency and Faulted)

Limit Load LEFM Weld Normal / Upset Emergency / Faulted Designation SF SF SF HI 88.0 41.9 --- H2 89.1 42.9 --- H3 50.5 24.7 4.2 (faulted)") H4 33.0 17.0 11.6 (upset)* , HS 50.3 26.1 --- H6 36.5 21.3 --- H7 39.5 22.6 ---

               "3 Indication #5, Uninspected area from 169.75* to 189.2*, and Indication #6

• Indications #34, #35, #36, Uninspected area from 349.82 to 9.40*, and Indication #1 I

Table 2-4. Calculated Flaw Lengths vs. Screening Criteria 1

 ~

Calculated Screening Criteria Flaw Length Allowable Flaw Length W eld (in) (in) Designation LimitLoad LEFM Limit Load LEFM H1 177 --- 501 --- H2 116 --- 498 --- H3 304 144 469 376 H4 362 79 460 310 H5 131 --- 450 --- H6 134 --- 422 -- I H7 74 --- 414 --- I l 12 _. l

    . ..      . . - . . . .                      . . - - -.       - . - - = . - - . - - .          - - . - . -      _               . - ._. ._-

GE Nuclear Energy GENE.523.Altu.n395 Ression i i

           ,                  2.1 Consideration of Additional Crack Growth To demonstrate the margin available in the core shroud welds, additional calculations were performed including an additional cycle of crack growth (total of two cycles beyond outage 10 UT results). Thus, calculations were performed by adding (2(2Aa) + U], where da is crack growth at each flaw end for one cycle, and U is the length uncertainty. Note

!. that this calculation is for the intent of demonstrating the margin available in the core shroud welds. This calculation also does not account for any new crack initiation. Tables 2-5 and 2-6 provide the results for these calculations. These results also indicate that the screening criteria and minimum required flaw evaluation safety factors are met with the additional operating cycle of crack growth. Some of the observed indications at welds H1, H3, H4, and H5 were combined for this evaluation due to the added crack l- growth and NDE uncertainty and due to the proximity criteria. Table 2-7 shows which indications were combined. I l L 1 l l-i 1 . 1 I.* 13

             - .  -.      -                         ~ - _        - .             ~.         . -      __ _         _               -   .--

GE Nuclear Energy GENE 321.A104 0893 Revision i l-Table 2-5. Flaw Evaluation Calculated Safety Factors With Crack Growth Assuming Two Operating Cycles (Required SF: 2.77 for Normal and Upset,1.39 for Emergency and Faulted)

  ,                                                                   Limit Load                                 LEFhl W eld                       Normal / Upset           i Emergency / Faulted Designation                            SF                          SF                            SF

!" HI 86.0 40.9 --- H2 88.2 42.5 --- H3 48.4 23.7 4.1 (faulted)") l H4 28.4 14.7 11.2 (upset)*

  .,                        HS                              49.2                      25.6                             ---

H6 36.1 21.1 ---

H7 39.1 22.4 ---

                                                                                                                                          \

(1) Indication #5, Uninspected area from 169.75* to 189.20*, and Indication #6

   ~

(2) Indications #34, #35, #36, Uninspected area from 349.82* to 9.40*, and Indication #1 Table 2-6. Calculated Flaw Lengths vs. Screening Criteria j With Crack Growth Assuming Two Operating Cycles Calculated Screening Criteria l

   .,                                                              Flaw Length              Allowable Flaw Length                         l W eld                       (in)                              (in)

Designation Limit Load LEFAI

    ~

Limit Load l LEF31 HI 186 -- 501 ---  ! H2 120 - 498 --- HD 315 147 469 376 H4 394 82 460 310

    ,                                        H5                137                --

450 --- H6 137 -- 422 --- l H7 78 -- 414 --- I4 ee

e GE Nuclear Energy GLVE-533 A104-0895 Resuien 1 a Table 2-7. Combined Indications for Two Operating Cycles HI Indication #1 and Uninspected area from 340.54' to 11.20' Indications #5, #6, and #7 i H3 Indications #3 and #4 Indication #5 and Uninspected area from 169.75' to 189.20' Indications #8 and #9 Indication #10 and Uninspected area from 352.97* to 11.20' and Indication #1 H4 Indications #2, #3, #4, and #5 Indications #6, #7, #8, #9, #10, #11, #12, and #13 Indications #19 and #20 Indications #21 and #22 Indications #23 and #24 Indications #27, #28, #29, #30, #31, and #32 Indications #34, #35, and #36 Indication #1 and Uninspected area from 349.82* to 9.40* H5 Indications #2 and #3 Indications #4. #5, #6, #7, #8, #9 and Uninspected area from 351.20* to 9.20* j 1 8 9 6 4 mm 15

_. . . . -. _ - - _ . _ . ~ . - _ - _ . _ _ _ . _ _ _ _ _ _ . - _ . . _ _ - - _ _ . . _ . _ _ _ _ - _ _ _ . _ l- '* ~ GENucle r Enero GENE.JU.AlN.0US Reshion !

          .              2.2              References 2-1.             BWR Core Shroud Distributed Ligament Length Computer Program, GE-NE-523-113 0894, Supplement 1, September 1994.

2-2. BWR Shroud Cracking Generic Safety Assessment, GE-NE-523-A107P-0794, Revision 1, Class III, August 1994. 2-3 Power Rerate Safety Analysis Report for Peach Bottom 2/3, NEDC-32230P, May 1993. se e d e em 4 em - k

    +

Mt a e

     .                                                                                    16

_ . . - . . ..--. - -- . ~ . - .- - - . - - - - - . - . l GENuclear Energy i e

• GENE 523.A104-0393 Rchion 1

(. . I

            ,            3.           

SUMMARY

AND CONCLUSIONS This report presents the screening criteria and flaw evaluation results for the core shroud l circumferential welds. The screening criteria was calculated using the up-to-date seismic !. and LOCA loads. UT inspection of the core shroud welds was performed during the 1995 fall outage (Outage 10). The evaluation assumes all UT detected indications are through-wall even though UT confirmed that they are only part through-wall. By meeting the screening cri:eria and exceeding the required safety factors using the flaw evaluation methodology, the ASME l Code Section XI safety margins are demonstrated to be satisfied. Both the screening criteria and flaw evaluation methods use linear elastic fracture

     ,                   mechanics (LEFM) and limit load concepts to determine acceptable through-wall indication lengths. The limiting flaw length based on either LEFM or limit load was used
     ~

for the screening criteria. For the Peach Bottom Unit 3 core shroud, only welds H3 and H4 were evaluated using LEFM. The screening criteria and flaw evaluation also use the ASME Code Section XI criteria for combining flaws based on the proximity ofindications. In addition, a second method for including the interaction between neighboring indication tips was considered for the LEFM allowable flaw size calculation. Results of the evaluation indicate that the screening criteria is satisfied at all weld locations. In addition, the flaw evaluation indicates safety factors wellin excess of the required safety factors. Thus, structural integrity over the next two year operating cycle is demonstrated. t I 9 ag 17

l GE Nuclear Energy GENE.331.A104 0893 o . Reshion 1 ? O 1 l APPENDIX A l l UT Inspection Reports for Welds H1 through H7 l I 1 5 l l s e e

e GE NuclearEnergy Peco Energy Peach Bottom 3R10 Shroud UTPro'act s 1CKSC September 1995 Shroud Weld H1 Indication Data TotalScan Length (Deg.) 304.10 TotalFlaw Length (Deg.) 14.56 TotalScan Length (In.) 583.83 TotalFlaw Length (In.) 27.95 Percentage of Weld Length Examined 84.5 Thickness (In.) 2.00 Percentage of Examined Weld Length Flawed 4.8 Circumference (In.) 691.15 Percentage of Total Weld Length Flawed 4.0 Inches perDegree 1.92 Indication Start End Length Length Max. Depth Max. Depth *A of Initiating Length Depth Number Azimuth Azimuth Degrees Inches Inches Pos. (Deg.) Thruwall Surface Transducer Transducer i 13.44 16.24 2.80 5.38 0.40 15.38 20.0 ID/Near 45' Shear 60*Long. 2 21.84 23.52 1.68 3.23 0.70 23.22 35.0 ID/Near 45'Sheat 60' Long. 3 38.20 39.88 1.6 C 3.23 0.36 39.02 18.0 IDINear 45' Shear 60' Long. 4 107.60 109.84 2.24 4.30 0.42 108.98 21.0 ID/Near 45' Shear 60'Long. 5 259.28 262.08 2.80 5.38 0.42 259.54 21.0 IDINeat 45' Shear 60'Long. 6 264.76 265.88 1.12 2.15 0.57 264.46 28.5 IDINeat 45'Sheat 60'Long.

                    '7         268.68 270.92          2.24      4.30        0.73      268.94        36.5      ID/Near   45' Shear 60'Long.
              *The deepest through wallIndicat]on sized.

Areas Not Examined by All3 Transducers 0* to 11.2',167.46' to 192.70' & 340.54' to 0* (Total of 55.90* Not Examined) l l Limitations: Core Spray Downcomers and Lifting Lugs l l I 1

                 .      . .. . . . . ..              .        j
                                                                      / C:i c 7i.,o
                                                            ,t                                                                 page z. or (z.

Resim 0 l

i . GE Nuclear Energy Poco Energy Peach Bottom 3R10 Shroud UTProject 1CKSC September 1995 Shroud Weld H2 Indication Data TotalScan Length (Deg.) 304.10 TotalFlaw Length (Deg.) 0.00 Total Scan Length (In.) $83.83 TotalFlaw Length (In.) 0.00 Percentage of Weld Length Examined 84.5 Thickness (In.) 1.00 Percentage of Examined Weld Length Flawed 0.0 Circumference (In.) 891.is Percentage of Total Wald Length Flawed 0.0 Inches per Degree 1.92 Indication Start End Length Length Max. Depth Max. Depth  % of Initiating Length Depth \ Number Azimuth Azimuth Degrees Inches Inches Pos. (Deg.) Thruwall Surface Transducer Transducer No RelevantIndications Recorded Areas Not Examined by AI!3 Transducers 1 0* to 11.4*,167.66* to 192.90* & 340.74* to 0* (Total of $5.90* Not Examined) l Limitations: Core Spray Downcomers and Lifting Lugs l l l l P.3'%Y ~ ~7 C:*~ ) di y CG G T'i5 Page 2- of /0 Revision 0

~ . 1 3

  *--~

GE NuclearEnergy Poco Energy Peach Bottom 3R10 Shroud UTProject 1CK5C September 1995 , Shroud Weld H3 Indication Data l TotalScan Length (Deg.) 322.32 TotalScan Length (In.) TotalFlaw Length (Deg.) 112.54 682.57 TotalFlaw Length (In.) 203.41 Percentage of Weld Length Examined 89.5 Percentage of Examined Weld Length Flawed Thickness (In.) 2.00 34.9 Percentage of Total WaldLength Flawed Circumference (In.) 650.67 31.3 Inches perDegree \ 1.61 ) Indication Start End Length Length Max. Depth Max. Depth  % of Initiating Length Depth Number Azimuth Azimuth Degrees Inches Inches Pos. (Deg.) Thruwall Surface Transducer Transducer 1 11.20 15.60 4.40 7.95 0.45 10.55 22.5 IDINear 45' Shear 60'Long. 2 54.20 62.45 8.25 14.91 0.72 57.75 36.0 IDINear 45' Shear 60'Long. 3 104.70 106.35 1.65 2.98 0.43 106.05 21.5 IDINeer 4 106.90 110.20 3.30 45' Shear 60'Long.  ! 5.96 0.40 10d.25 20.0 ID/Near

              '5                                                                                                           45' Shear    60' Long.      '

144.05 169.45 25.40 45.91 0.85 163.10 42.5 ID/Near 6 203.21 232.65 29.44 45' Shear 60'Long. 53.21 0.78 224.50 39.0 IDINear 7, 240.92 250.32 9.40 45' Shear 60*Long. 16.99 0.64 244.54 32.0 IDINear 5 298.68 309.33 10.85 45' Sheat 60'Long. 19.25 0.60 302.30 30.0 ID/Near

              *9        310.88 325.32        14.44 45' Shear 60'Long.

26.10 0.85 323.90 42.5 ID/Near "10 348.72 354.33 45' Shear 60*Long. 5.61 10.14 0.65 350.10 32.5 ID/Near 45' Shear 60'Long.

         'The deepest through.wallIndication sized.
         " Length sizing ofIndication nio is restricted by the limitation of the core spray downcomer Areas Not Examined by All3 Transducers O' to 11.2',169.75* to 189.2* & 352.97* to 0* (Total of 37.68* Not Examined)

Limitations: Core Spray Downcomers and Lifting Lugs l l "U5555555Idd . . r DCT 07'95 l l l l

                                         .                                                                                                              l 1

Page E of 5 Revision 1

                                                                                                                                                                          = . ,

I i m W GENuclearEnergy J l Poco Energy , ? Peach Bottom 3R10 Shroud UTProject 1CKSC September 1995 l Shroud Weld H4 Indication Data ' Total Scan Length (Oeg) 321.04 TotalFlaw Lenge (Deg) 103.30 Total Scan Length (In) 580.25 TotalFlow Lenge (In) 188.71 t l l Percentage of Weld Length Esamined 89.2 l Thickness (In) 2.00 1 Percentage of Examined Weld Length Flawed 32.2 Circumference (In) 850.87 Vercentage of Total Weld Length Flawed 28.7 laches per Degree 1.81 Indication Start End Length Length Max. Depth htax. Depth  % of Initiating Length Depe Number himuth himuth Degrees Inches Side of Inches Pos. (Deg.) Thruwell Surface Transducer Transducer Weld 1 10.32 11.44 1.12 2.02 " " " ID/Near 45' Shear " Lower 2 23.70 28.50 2.80 5.08 " " " ID/Near " 45' Shear Upper 3 24.78 25.88 1.12 2.02 " " " ID/Near " 45' Shear Lower 4 27.00 28.88 1.88 3.04 " " " ID/Near 45' Shear " Lower i 8 28.18 29.30 1.12 2.02 " " " lDINest " 45' Shear Upper 8 38.02 37.14 1.12 2.02 " " " IDINear " 45' Shear Upper 7 42.00 48.38 3.38 8.07 " " " lDINear " 45' Shear Lower 8 47.88 51.88 3.92 7.09 " " " ID/Near " 45' Shear Upper 9 49.28 84.32 8.04 9.11 " " " fDINear " 45' Shear Lower 10 58.32 87.88 2.24 4.05 " " " ID/Near " 45' Shear Lower

            .           11      82.10     87.70     8.80       10.12         0.13         86.42            8.5     ID/Near    45* Shear 80*Long.        Upper
      *W                12      83.18     84.28     1.12        102           "  .

IDINear 45* Shear " Lower 13 72.08 73.18 1.12 2.02 " " " lD/Near " 45' Shear Upper ' 14 83.70 84,82 1.12 2.02 " " " ID/Near 45' Shear " Upper il 98.82 99.32 2.80 5.08 " " " I ID/Near 45' Shear " Lower il 113.28 114.28 1.00 1.81 " " "  ! ID/Near 45* Shear " Upper 17 124.34 125.48 1.12 2.02 " " " j lO/Near 45' Shear- " Upper

                       *18      135.M 150.38        15.00      27.11        >$0%         140.18           >$0%     @Near      45' Shear 80'Long.

l Lower l 19 201.02 205.38 4.38 7.88 " " " IDINear 45* Shear " Upper l 20 202.08 204.32 2.24 4.05 " " " ID/Near 45' Shear " Lower 21 210.98 213.22 2.24 4.05 " " " IDINear 45' Shear " Upper 22 218.02 218.82 2.80 8.06 " " " ID/Near 45' Shear " Upper  ! 2J 230.40 232.08 1.88 3.04 " " " ID/Near 45* Shear " Lower 24 233.70 235.38 1.88 3.04 " " " lD/Near 45' S hear " Upper 25 244.84 247.08 2.24 4.05 " " " 1DINear 45* Shear " Lower 28 283.70 285.38 1.88 3.04 " " " IDINear 45* Shear " Upper 27 289.98 293.78 3.80 8.87 " " " 10/Near 45' Shear " Lower

                     ' 28       294.32 295.44        1.12       2.02          "            "                "

ID/Near 45' Shear " Lower 29 298.50 297.62 1.12 2.02 " " " ID/Near 45' Shear " Upper 30 298.88 297.88 1.12 2.02 " " " ID/Near 45' Shear " Lower 31 298.24 301.80 3.38 8.07 " " " IDINear 45' Shear " Lower 32 308.38 308.78 2.88 4.84 " " " IDINear 45' Shear " Lower 33 318.28 319.40 1.12 2.02 " " " IDINear 45' Shear " Lower 34 324.88 339.18 14.30 25.85 0.14 338.38 7.0 ID/Near 45' Shear 80*Long. Lower 35 325.38 327.06 1.88 3.04 " " " ID/Near 45' Shear " Upper 38 340.30 341.98 1.88 3.04 " " " IDINear 45* Shear " Lower 9he deepest through wallIndication sized. Upper 34.s8 (Deg.)

                 " Thru wall dimension not obtained due to flaws being below our sizing threshvid. (0.10*)                         Lower    88.82    (Deg.)

Without Overlapping 93.78 (Deg.) Areas Not Esamined by All3 Transducers 0' to 9.4',170.02* to 189.40* A 349.82* to O'(Total of 38.98 Not Examined) Upper 82.32 ! (In.) I

                                                                                                                      .            Lower    124.39   (In.)
      %..        Limitations: Core Sprey Downcomers and Lifting Lups                                                 Wtthout Over1apping    159.50   (in.)
                "                                                                                                                                Paselefd
                         ..l:.'.h.%,,,2,, W JJ,
                                              .                       .            ocro 7'e5                                                         ""**"

_ _ . ~ . - . __ . _ _ - .- -. - - . - - - . - - - i \ t .. z e GENuclearEnergy Peco Energy Peach Bottom 3R10 Shroud UTProject 1CKSC September 1995 i Shroud Weld H5 Indication Data TotalScan Length (Deg.) 326.80 TotalFlaw Length (Deg.) 24.64 TotalScan Length (In.) 590.66 TotalFlaw Length (In.) 44.53 i Percentage of WeldLength Examined 90.8 Percentage of Examined Weld Length Flawed Thickness (In.) 2.00 7.5 Percentage of TotalWeldLength Flawed Circumference (In.) 650.67 6.8 Inches per Degree 1.81 Indication Start End Length Length Max. Depth Max. Depth  % of Initiating Length Depth Number Asimuth Azimuth Degrees Inches Inches Pos. (Deg.) Thruwall Surface Transducer Transducer 1 141.52 144.88 3.36 6.07 0.11 142.34 5.5 2 319.34 324.38 5.04 lD/Near 45' Shear 60'Long. 9.11 0.20 322.34 10.0 ID/Near

                     *3                                                                                                                45' Shear   60' Long.

325.38 328.18 2.80 5.06 0.23 326.14 11.5 ID/Near 4 333.78 336.58 45' Shear 60'Long. 2.80 5.06 0.14 334.54 7.0 IDINear 5 338.26 339.38 45' Shear 60'Long. 1.12 2.02 0.20 338.46 10.0 ID/Near 6 336.26 338.50 45' Shear 60*Long. 2.24 4.05 0.11 337.02 5.5 7 339.62 341.86 2.24 ID/Near 45' Sheat 60'Long. 4.05 0.11 340.38 5.5 ID/Near J

                   .&          344.10 346.90      2.80 45' Shear 60'Long.

5.06 0.18 345.42 9.0 9 348.02 350.26 2.24 ID/ Neat 45' Shear 60'Long. 4.05 0.10 348.22 5.0 ID/Near 45' Shear 60'Long.

       *** *The deepest through-wallIndication sized.

Areas Not Examined by A!! 3 Transducers Areas Not Examined: 0* to 9.20',174.20* to 189.40' & 351.20' to 0* (Total of 33.20* Not Examined) Limitations: Core Spray Downcomers and Lifting iugs 1 PageE-ofl2-- dc - Ol.'T. f.' . -'O% Revision 0 7

i l

 ..     , .                                                                                                                                             I i

e , e

    .i l
     ,                        GE NuclearEnergy Poco Energy Peach Bottom 3R10 Shroud tJTProject 1CKSC September 1995 1

Shroud Weld H6 Indication Data 1 TotalScan Length (Deg.) 288.52 TotalFlaw Length (Deg.) 0.00 TotalScan Length (In.) 606.08 TotalFlaw Length (In.) 0.00 Percentage of Weld Length Examined 80.1 Thickness (In.) 2.00 t Percentage of Examined Weld Length Flawed 0.0 Percentage of Total Weld Length Flawed Circumference (In.) 831.46 el l 0.0 Inches per Degree 1.75 i L  ! i Indication Start End Length Length Max. Depth Max. Depth  % of Initiatitv. Length Depth Number Azimuth Azimuth Degrees Inches Inches Pos. (Deg.) Thruwall Surface Transducer Transducer l No RelevantIndications Recorded Areas Not Examined by AII3 Transducers \ 0* to 9.2*,166.96* to 219.20' & 349.96* to 0* (Totalof 71.48* Not Examined) ' i Limitations: Core Spiny Downcomers and Lifting Lugs l l I Mf(fs

                   ,;                                                         C ; '. 7 i5 i

l l t Page 2-of Revision 0

b i . - g ., .t

• GE Nuclear Energy Poco Energy Peach Bottom 3R10 Shroud tJTProject 1CKSC September 1995

} 1 Shroud Weld H7 Indication Data 1 d TotalScan Length (Deg.) 322.64 TotalFlaw Length (Deg.) 0.00 TotalScan Length (In.) 666.93 TotalFlaw Length (In.) 0.00 ^ 1 Percentage of Weld Length Examined 89.6 Thickness (In.) 2.00 i Percentage of Examined Weld Length Flawed 0.0 i Circumference (In.) 631.46 l Percentage of Total Weld Length Flawed 0.0 Inches per Degree 1 1.75 l 1 i Indication Start End Length Length Max. Depth Max. Depth  % of Initssting Length Depth Number Azimuth Azimuth Degrees Inches Inches Pos. (Deg.) Thruwall Surface Transducer Transducer i No RelevantIndications Recorded Areas Not Examined by All3 Transducers 0* to 9.4*,170.92* to 189.40* A 350.52* to 0* (Totalof 37.36' Not Examined) i Limitations: Core Spray Downcomers and Lifting Lugs , GN l l I 1 RI t.  : ; ?Z.T.E !?=:T..~S?Md

u :.. c . w .

OCT 07 T5 9 Page E- of (O Revision 0

   ~ _ _ _ . -                   _    _ _ _ . _ _ _ _ - -                _ _. . _ _ _ _ _ __ _ _ _ - .      .. _ ___ _ ___ _

• l l

          ',          f9,,a nog),,

I E S UNITED STATES

        ?      .

1 *i NUCLEAR REGULATORY COMMISSION

        '[-           %....+,/                                   WASHINGTON, D.C. 20555-0001 February 6, 1995 i

Mr. George A. Hunger, Jr. Director-Licensing, MC 62A-1 PECO Energy Company i Nuclear Group Headquarters  ! Correspondence Control Desk P.O. Box Ho. 195 l Wayne, PA 19087-0195 l

SUBJECT:

GENERIC LETTER (GL) 94-03, "INTERGRANULAR STRESS CORROSION CRACKING l OF CORE SHROUDS IN BWRs," PEACH BOTTOM ATOMIC POWER STATION, UNIT l NOS. 2 AND 3, (TAC NOS M90105 AND M90106) l

Dear Mr. Hunger:

By letter dated August 24, 1994, the PECO Energy Company (PEco) provided its ' response to Generic Letter (GL) 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs," for the Peach Bottom Atomic Power Station, Units 2 and 3. The NRC staff requested in GL 94-03 that licensee's teko the ! following actions with respect to their core shrouds: (1) inspect timir core shrouds in their BWR plants no later than the next refueling ondtge; (2) perform materials-related and plant-specific consequance safety analyses with respect to their core shrouds; (3) develop core shrord inspection plans which address inspection of all core shroud welds and titch takes into account the latest available inspection technology; (4) develcp plans for evaluation and/or repair of their core shrouds; and (b) work closely with the BWR Owners Group with respect to addressing intergranular stress corrosion cracking of BWR internals. The NRC staff requested that licensee's submit, under oath or affirmation, the following information in response to GL 94-03 within 30 days of the date of issuance: (1) a schedule for inspection of their core shrouds; (2) a safety analysis, including a plant-specific safety analysis as appropriate, which supports continued operation of the facility until inspections are conducted; (3) a drawing (s) of the core shroud configurations; and (4) a history of shroud inspections completed to date. The NRC staff also requested that licensee's submit, under oath or affirmation, no later than 3 months prior to performing their core shroud inspections, their scope for inspection of their core shrouds and their plans for evaluating and/or repairing their core shrouds based on their inspection results. The NRC staff further requested licensee's to submit, under oath or affirmation, their core shroud inspection results and flaw evaluation within 30 days of completing their shroud examinations. Based on the staff's review of PEco's August 24, 1994, response to GL 94-03, and in regard to the information that was requested to be submitted within 30 days of the date of issuance of the Gl the staff concludes that PECo has provided the operational, fabrication and materials related information requested for both the Peach Bottom Units 2 and 3. l l 00 6 ff[ _. _

j .,- 1 ! i

.- 1 G. Hunger, Jr. .i 4 The staff notes that PEco has previously examined the Peach Bottom Unit 3

! (PBAPS 3) core shroud during refueling outage (RFO) 3R09. Based on the j results of the materials-based structural analysis of the PBAPS 3 core shroud, i the staff concludes that the structural margins for the PBAPS 3 core shroud  ! l will be maintained during the current PBAPS 3 operating cycle (Unit 3, Cycle l J 10). The staff therefore concludes that the results of the licensee's ! materials-based structural analysis are sufficient to justify continued safe I operation for the remainder of the current PBAPS 3 operating cycle without i necessitating a detailed consequence safety analysis or a modification of the  ! PBAPS 3 core shroud. The staff's evaluation of PEco's GL 94-03 reponse for ' ! Unit 3 is provided as Enclosure 1. 1 Per the reporting requirements of GL 94-03, the licensee is reminded that, for i inspection scope and shroud evaluation / repair information that has been i requested but not yet been submitted (i.e., PBAPS Unit 3), the inspection ! scope and evaluation / repair scope information should be submitted within 3 i months of performing their scheduled core shroud inspections. W. M90106 will i remain open for Peach Bottom Unit 3 pending submittal of the shread l inspection / repair plans. . e s The staff also notes that PEco has recently completed the Peach Bottom ' Unit 2 (PBAPS 2) core shroud examinations, which were performed during the recently completed refueling outage 2R010, and which were performed per the actions requested by GL 94-03. The staff has received the November 7, 1994 submittal containing the results and evaluation of the PBAPS 2 core shroud examinations which were performed during RF0 2R10. The results of the PBAPS 2 shroud inspections indicate that the cracks in the PBAPS 7. core shroud are bounded by those recorded for PBAPS 3, and are therefore acceptable for service during the next operating cycle (Operating Cycle 11). The staff's evaluation of PEco's GL 94-03 reponse for Unit 2 is provided as Enclosure 2. Staff action for PECO's response to GL 94-03 for Unit 2 is completed and TAC M90105 is closed. Sinc ely, G /} JosephT.Shea,ProjectManager Project Directorate 1-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation Docket Nos. 50-277/278

Enclosures:

As stated - cc w/encls: See next page l Il

         .          Mr. George A. Hunger, Jr.                                           Peach Bottom Atomic Power Station,
       ,            PECO Energy Company                                                   Units 2 and 3

• 8 cc:

J. W. Durham, Sr., Esquire Mr. Rich R. Janati, Chief Sr. V.P. & General Counsel Division of Nuclear Safety PECO Energy Company Pennsylvania Department of 2301 Market Street, S26-1 Environmental Resources Philadelohta, Pennsylvania 19101 P. O. Box 8469 Harrisburg, Pennsylvania 17105-8469 PECO Energy Company 1 ATTN: Mr. G. R. Rainey, Vice President Board of Supervisors Peach Bottom Atomic Power Station Peach Bottom Township Route 1, Box 203 R. D. #1 l Delta, Pennsylvania 17314 Delta, Pennsylvania 17314 PECO Energy Company Public Service Commission of Maryland ATTN: Regulatory Engineer, A4-SS Engineering Division

Peach Bottom Atomic Power Station Chief Engineer Route 1, Box 208 6 St. Paul Centre Delta, Pennsylvania 17314 Baltimore, MD 21202-6806 Resident Inspector Mr. Richard McLean U.S. Nuclear Regulatory Commission Power Plant and Environmental Peach Bottom Atomic Power Station Review Division P.O. Box 399 Department of Natural Resources Delta, Pennsylvania 17314 B-3, Tawes State Office Building l Regional Administrator, Region I U.S. Nuclear Regulatory Commission Mr. John Doering, Chairman 475 Allendale Road Nuclear Review Board King of Prussia, Pennsylvania 19406 PECO Energy Com>any l

965 Chesterbroot Boulevard Mr. Roland Fletcher Mail Code 63C-5 Department of Environment Wayne, Pennsylvania 190 % 201 West Preston Street Baltimore, Maryland 21201 Dr. Judith Johnsrud Natic.1al Energy Committee A. F. Kirby, Ill Sierra Club External Operations - Nuclear 433 Orlando Avenue Delmarva Power & Light Company State College, PA 16803 P.O. Box 231 ! Wilmington, DE 19899 Mr. Richard Ochs Maryland Safe Energy Coalition P.O. Box 33111 Baltimore, MD 21218 l i l 1

i ', /'pa ny~%, l E  % UNITED STATES 5

• 1 NUCLEAR REGULATORY COMMISSION

         '.<                                                                                                                 l g*****                                             WASHINGTON, D.C. 20566 4 001

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RESPONSE TO GENERIC LETTER 94-03

\ PEACH BOTTOM ATOMIC POWER STATION. UNIT 3 l PECO ENERGY COMPANY DOCKET NO. 50-277 1.0 BACKGROUNO The core shroud in a Boiling Water Reactor (BWR) is a stainless steel cylindrical component within the reactor pressure vessel (RPV) that surrounds the reactor core. The core shroud serves as a partition between feedwater in the reactor vessel's downcomer annulus region and the cooling water flowing up through the reactor core. In addition, the core shroud provides a refloodable  ; volume for safe shutdown cooling and laterally supports the fuel assemblies to maintain control rod insertion geometry during operational transients and accidents. In 1990, crack indications were observed at core shroud welds located in the beltline region of an overseas BWR. This reactor had completed approximately 190 months of power operation before discovery of the cracks. As a result of this discovery, General Electric Company (GE), the reactor vendor, issued Rapid Information Communication Services Information Letter (RICSIL) 054,

                                   " Core Support Shroud Crack Indications," on October 3,1990, to all owners of GE BWRs.             The'RICSIL summarized the cracking found in the overseas reactor and recommended that at the next refueling outage plants with high-carbon-type 304 stainless steel shrouds perform a visual examination of the accessible areas of the seam welds and associated heat-affected zone (HAZ) on the inside and outside surfaces of the shroud.

Subsequently, a number of domestic BWR licensees performed visual examinations of their core shrouds in accordance with the recommendations in GE RICSIL 054 or in GE Services Information Letter (SIL) 572, which was issued in late 1993 to incorporate domestic inspection experience. Of the inspections performed to date, significant cracking was reported at several plants. The combined industry experience from these plants indicates that both axial and circumferential cracking can occur in the core shrouds of GE designed BWRs. On July 25, 1994, the NRC issued Generic Letter (GL) 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors," to all BWR licensees (with the exception of Big Rock Point, which does not have a core shroud) to address the potential for cracking in their core shrouds. GL 94-03 requested BWR licensees to take the following actions with respect to their core shrouds: inspect their core shrouds no later than the next scheduled refueling outage; Enclosure 1 g[g o noc~ (6f y-

e 1 1 perform a safety analysis supporting continued operation of the facility until the inspections are conducted; develop an inspection plan which addresses inspections of all shroud welds, and which delineates the examination methods to be esed for the inspections of the shroud, taking into consideration the best industry technology and inspection experience to date on the subject; l develop plans for evaluation and/or repair of the core shroud; and l - work closely with the BWROG on coordination of inspections, evaluations, t and repair options for all BWR internals susceptible to intergranular l stress corrosion cracking. l l The PECO Energy Company (PECo), the licensee for the Peach Bottom Atomic Power Station Unit 3 (PBAPS 3), responded to GL 94-03 on August 24, 1994 (Reference i 1). Part of the licensee's response included PEco's inspection scope for the l 1 planned re-inspections of the PBAPS 3 core shroud, which have been scheduled for refueling outage (RFO) 3R10 in the fall of 1995. The licensee completed an inspection of the PBAPS 3 core shroud during the previous RF0 in the fall r of 1993. The General Electric Nuclear Energy Division formally submitted the examination results and assessment of core shroud structural integrity to the l NRC by letter dated December 3, 1993 (Reference 2). PEco amended the results and assessment by letter dated March 14, 1994 (Reference 3). 2.0 STAFF'S EVALUATION OF THE LICENSEE'S RESPONSE TO GL 94-03 PEco completed a limited visual inspection of the PBAPS 3 core shroud during the 3R9 RF0 in the fall of 1993. The licensee has planned a more comprehensive inspection of the PBAPS 3 core shroud for the next RF0, scheduled for the fall of 1995. 2.1 Suscentibility of the PBAPS 3 Core Shroud to IGSCC The core shroud cracks which are the subject of GL 94-03, result from intergranular stress corrosion cracking (IGSCC) which is most often associated with sensitized material near the component welds. IGSCC is a time-dependent phenomena requiring a susceptible material, a corrosive environment, and a tensile stress within the material. Industry experience has shown that austenitic stainless steels with low carbon content are less susceptible to IGSCC than stainless steels with higher carbon content. BWR core shrouds are constructed from either type 304 or 304L stainless steel. Type 304L stainless steel has a lower carbon content that type 304 stainless steel. During the shroud fabrication process when the l sections of the core shroud are welded together, the heating of the material ! adjacent to the weld metal sensitizes the material. Sensitization involves carbon diffusion out of solution forming carbides at grain boundaries upon moderate heating. The formation of carbides at the grain boundaries depletes

l' ] , 3-i j the chromi'an in the adjacent material. Since the corrosion resistance of j stainless steel is provided by the presence of chromium in the material, the uea adjacent to the grain boundary depleted of chromium is thereby i susceptible to corrosion. j Increased material resistance to IGSCC will result if the material, carbon content is kept below 0.035%, as specified for type 304L grade l i Currently available inspection data indicate that shrouds fabricated with forged ring segments are more resistant to IGSCC than rings constructed from i welded plate sections. The current understanding for this difference is ! related to the surface condition resulting from the two shroud fabrication ! processes. Welded shroud rings are constructed by welding together arcs machined from rolled plate. This process exposes the short transverse i direction in the material to the reactor coolant. Elongated grains and 3 stringers in the material exposed to the reactor coolant environment are j believed to accelerate the initiation of IGSCC. l i Water chemistry also plays an important role in regard to IGSCC ) susceptibility. Industry experience has shown that plants which have operated i with a history of high reactor coolant conductivity have been more, susceptible to IGSCC than plants which have operated with lower conductivities , J Furthermore, industry experience has shown that reactor coolant systems RCSs) { which have been operated at highly positive, electrochemical potentials ((ECPs) havebeenmorgsusceptibletoIGSCCthanRCSsthathavebeenoperatedatmore l negative ECPs . The industry has made a considerable effort to improve water i chemistry at nuclear facilities over the past 10 years. Industry initiatives ! have included the introduction of hydrogen water chemistry as a means of j lowering ECPs i.e., making the ECPs more negative) in the RCS. The , effectiveness o(f hydrogen water chemistry in reducing the susce)tibility of ! core shrouds to IGSCC initiation has not been fully evaluated; lowever, its effectiveness in reducing IGSCC in recirculation system piping has been ! demonstrated. l Welding processes can introduce high residual stresses in the material at the 1 i l $ ' Conductivity is a measure of the anionic and cationic content of j liquids. -As a reference, the conductivity of pure water is ~0.05 s/cm. 4 Reactor coolants with conductivities below 0.20 us/cm are considered to be t relatively ion free; reactor coolants with conductivities above 0.30 s/cm are considered to have a relatively high ion content. 2 The electrochemical potential (ECP) is a measure of a material's 4 susce)tibility to corrosion. In the absence of an externally applied current, j and tierefore, for reactor internals in the RCS, the electrochemical potential

is equal to the open circuit potential of the material. Industry experience i has shown that crack growth rates in reactor internals are low when the

! ECP s --0.230 volts. l j 4

- )

i . . A i I weld joint. The high stresses result from thermal contraction of the weld metal during cooling. A higher residual tensile weld stress will increase the ' material's susceptibility to IGSCC. Although weld stresses are not easily quantified, previous investigation into weld stresses indicate that tensile i stresses on the weld surface may be as high as the yield stress of the material. The stress decreases to compressive levels in the center of the j welded section. i PECo has reviewed the materials, fabrication and operational histories of the i PBAPS 3 core shroud and has submitted this information to the staff in their { response to Gl. 94-03. The PBAPS 3 plant-specific susceptibility factors are l summarized below: I The shroud support, top guide support, and core support plate rings are i fabricated from two welded 304 stainless steel, forged ring segments, with c arbon contents of ~0.030%. The shroud shell region was fabricated by welding rolled 304 stainless steel plates together. The carbon content of the PBAPS 3 shroud plates are in the range of 0.050 - 0.065%. Welding of the shroud plates and rings for circumferential welds H1 - H6 l was accomplished by submerged arc welding using ER308 filler metal. Welding of the bi-metallic weld, H7, was accomplished by gas metal arc l j { welding using filler metal 82. Weld residus1 stress levels resulting  ;

from these fabrication processes are high.

i PBAPS 3 operated at high reactor coolant ionic content levels during the

initial years of operation. The initial five year average coolant 1 i

conductivity for PBAPS 3 was 0.695 pS/cm, which is considerably higher '

than the average for other U.S. BWRs (where the conductivities range i

1 from ~0.123 pS/cm to 0.717 pS/ce, and average - 0.340 pS/ca). l PBAPS 3 has operated for 11 cumulative years at full power, which is slightly above the median for U.S. BWRs (range is 3.7 years - 17.8 years, with a median of 10.8 years).

A review of the plant-specific factors which increase the potential for IGSCC

in BWR core shrouds reveals that PBAPS 3 initially operated at high reactor

coolant conductivity during the first five cycles of operation. In addition, i

the carbon content of the taaterial which comprises the PBAPS 3 core shroud is  ;

! relatively high. On these bases, the Boiling Water Reactor Vessels & l j Internals Project (BWRVIP) has classified the PBAPS 3 core shroud as a ' susceptible Category "C" shroud. The staff has also determined that the PBAPS l 3 core shroud is susceptible to IGSCC, and therefore concludes that the BWRVIP's susceptibility assessment is acceptable. This conclusion is ] supported by the identification of moderate cracking during the previous core shroud inspection. This is discussed further in the following section. I i i l

,4 i    ..

i '.' I J 2.2 Insoection of the Peach Bottom Unit 3 Core Shroud a PEco inspected the PBAPS 3 core shroud during RF0 3R9 in the fall of 1993. i The staff previously reviewed the licensee's evaluation of the PBAPS 3 core i shroud and determined that the licensee's assessment justified continued operation of PBAPS 3 for the current operating cycle (Operating Cycle 10). ! The staff's assessments of the licensee's inspection scope and flaw evaluation 1 are provided in References 4 and 5 listed under Section 5.0 of this Safety Evaluation (SE). The following is a description and staff assessment of the licensee's core shroud inspection, j 2.2.1 Inoection Scone and Results for Core Shroud Examinations I 1 The intpections completed during RF0 3R9 were done in accordance with

!             recommendations of SIL-572, Revision 1. The scope of the inspections included 1

examination via enhanced VT-1 methods. The licensee initially completed a partial examination of the core shroud circumferential welds. Their original inspection sco)e required enhanced VT-1 examinations at eight (8) cell locations of tie H1, H2, H3, H4, and H5 welds. The licensee expanded the inspection scope after discovering indications at the H3 and H4 celds. The expanded scope included the following examinations: 100% enhanced VT-1 from the inside diameter (ID) of the H3 and H4

welds; i -

100% enhanced VT-1 of accessible areas of weld H4 on the outside 2 diameter (00); i . enhanced VT-1 examinations of the H3 weld from the 00 in areas where ! cracking was not indicated on the ID; an enhanced VT-1 examination of the H3 weld from the OD in areas 4 where cracking was indicated on the ID; enhanced VT-1 examinations at six (6) locations of the H6 weld; enhanced VT-1 examinations at two (2) locations of the respective H7 and H8 welds; enhanced VT-1 examination of one (1) vertical weld between the H3 and H4 welds; and

                      .. enhanced VT-1 examination of the of the mid-shroud plates.

j The licensee's VT-1 examinations identified a large (~105 inch) crack in the i H3 weld (the weld joining the top guide support ring to the upper mid-shroud shell). Less extensive cracking was also found at the H4 weld (< 30 inches total). Minor cracking was determined to exist at weld H1 and at one of the vertical shroud welds, t 2.2.2 Evaluation of the Peach Bottom 3 Core Shroud Insoection Results PEco's evaluation and disposition of the inspection data was the basis for

justifying operation of the PBAPS 3 Unit during the current operating cycle

(Cycle 10). PEco issued a preliminary draft on the Peach Bottom Unit 3 core shroud flaw evaluation during the PEco/NRC meeting of November 3,1993, at Rockville, Maryland. PEco formally submitted this flaw evaluation to the

staff on December 3,1993 (Reference 2), and amended it on March 14, 1994 (Reference 3). The licensee's flaw evaluation was performed in accordance with the methods found in General Electric (GE) Document GENE-523-141-1093,

         " Evaluation and Screening criteria for the Peach Bottom Unit 3 Shroud Indications," Rev. 0 (Reference 2) and Rev. 1 (Reference 3). The licensee's submittal included the results of the PBAPS 3 core shroud inspections performed during the previous RFO.

Flaw evaluations of the PBAPS 3 shroud were performed in accordance with the structural margin criteria found in Sectirsn XI of the ASME Code. Evaluations of the indications of the PBAPS 3 core throud, which included adjustments to account for crack proximities, crack growth and non-destructive examination uncertainties, indicated that the PBAPS 3 core shroud would maintain sufficient structural integrity for the current operating cycle (Operating Cycle 10). 2.2.3 Staff Assessumnt of the Peach Bottom Unit 3 Inspection and Evaluation The staff concluded (References 4 and 5), after reviewing PEco's inspection scope for the VT-1 examinations, that the inspection scope was sufficient to ascertain the condition of the PBAPS 3 shroud. The staff also concluded (References 4 and 5) that the licensee's flaw evaluation method was acceptable and that PBAPS 3 core shroud would meet structural margin requirements during the current operating cycle. PEco is required by GL 94-03 to submit its inspection scope for re-inspection of the PBAPS 3 core shroud 90 days prior to entering the fall 1995 RF0.

3.0 CONCLUSION

S Based on a review of the PBAPS 3 core shroud materials, fabrication processes and operating history, the staff concludes that the licensee's core shroud is susceptil,le to IGSCC. PEco completed an examination of the PBAPS 3 core shroud during RF0 3R9. The licensee's assessment of identified weld cracking indicates that the PBAPS 3 core shroud will maintain sufficient structural margins throughout the current operating cycle. The staff concluded that the licensee's flaw evaluation of the PBAPS 3 core shroud was acceptable and justified operation of the PBAPS 3 reactor for the current operating cycle (References 4 and 5). 4.0 OUTSTANDING ISSUES / FUTURE ACTIONS In accordance with the reporting requirements of GL 94-03, the licensee shall submit to the NRC, no later than 3 months prior to performing the core shroud i inspections, both the inspection plan and the licensee's plans for evaluating and/or repairing of the shroud based on the inspection results. In addition, , results should be provided to the NRC within 30 days from the completion of l 3 I

4 l the inspection. If the licensee identifies any core shroud cracking requiring an analysis per the ASME code, details of such evaluations must also be submitted to the NRC for review. It should be noted that the industry is currently encountering difficulties l performing comprehensive inspections of lower shroud welds and /or lower vessel regions due to NDE equipment accessibility problems. The staff urges licensees to work with the members of the EPRI NDE Center in order to develop l ' improved tooling for inspections of shroud welds and lower vessel regions which are highly obstructed. Should improved inspections techniques become I available, the staff recomendation is for licensee's to re-inspect the lower shroud welds at the earliest opportunity.

                                                                                                   \

At present, the NRC has not approved the inspection guidelines proposed by the l BWRVIP. Considerable differences remain with regard to the recommended scope of core shroud inspections. The staff cautions the licensee against modifying their plans according to BWRVIP recommendations which have not undergone review and approval by the NRC. The staff's current position with regard to the scope of inspections is a recomendation for the inspection of 100% of the accessible core shroud welds. Should the licent.ee opt to install a preemptive repair in lieu of performing a comprehensive core shroud inspection the only required inspection is that mandated in the staff approval of the repair option.

5.0 REFERENCES

1. Letter from G. A. Hunger, Jr., Director of Licensing, PECO Energy j Company, to the U.S. Nuclear Regulatory Commission fonvarding the " Peach Bottom Atomic Power Station, Units 2 and 3, Limerick Generating Station Units 1 and 2 Response to Generic Letter 94-03, 'Intergranular Stress I Corrosion Cracking of Core Shroud in Boiling Water Reactors,'" dated 1 August 24, 1994.

l

2. Letter from M. L. Herrera and H. Mehta, General Electric Nuclear Energy, i to the U.S. Nuclear Regulatory Commission forwarding the General Electric " Evaluation and Screening Criteria for the Peach Bottom Unit-3 Shrou& Indications," Rev. O, (GENE-523-141-1093) dated December 3, 1993.

3. Letter from G. A. Hunger, Jr., Director of Licensing, PECO Energy Company, to the U.S. Nuclear Regulatory Commission forwarding the General Electric " Evaluation and Screening Criteria for the Peach Bottom Unit-3 Shroud Indications," Rev. 1, (GENE-523-141-1093) dated March 14, 1994.

l l l l l

1

4. NRC internal memorandum from Jack R. Strosnider, Chief, Naterials and Chemical Engineering Branch, Division of Engineering, to Larry E.

'                      Nicholson, Acting Director, Project Directorate, I-2, Division of Reactor Projects I/II, forwarding staff's " Evaluation of Peach Bottom Shroud Cracks," dated November 9, 1993.

2 a 5. Stephen Dembek, Project Manager, Project Directorate I-2, Division of Reactor Projects - I/II, Office of Nuclear Reactor Regulation, issuance of " Meeting Summary, Evaluation of Core Shroud Indications at Peach Bottom, Unit 3 (TAC No. M88099)," dated December 2, 1993. Principal Contributor: J. Hedoff Date: February 6, 1995

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        -a ($e                .?x i                                S                                         UNITED STATES

• 5 "

    .-                                                       NUCLEAR REGULATORY COMMISSION

) g..... WASHINGTON, D.C. 2055H001 L j SAFETY EVALUATION BY THE OFFICF 0F NUCLEAR REACTOR REGULATION

RESPONSE TO GENERIC LETTER 94-03 j PEACH BOTTOM ATOMIC POWER STATION. UNIT 2 PECO ENEfLGY COMPANY 1

{ DOCKET NO. 50-277 I

1.0 BACKGROUND

! The core shroud in a Boiling Water Reactor (BWR) is a stainless steel ! cylindrical component within the reactor pressure vessel (RPV) that surrounds the reactor core. The core shroud serve: as a partition between feedwater in the reactor vessel's downcomer annulus region and the cooling water flowing up i' through the reactor core. In addition, the core shroud provides a refloodable volume for safe shutdown cooling and laterally supports the fuel assemblies to maintain control rod insertion geometry during operational transients and g accidents. 1 i In 1990, crack indications were observed at core shroud welds located in the I l beltline region of an overseas NR. This reactor had completed approximately i 190 months of power operation before discovery of the cracks. As a result of this discovery, General Electric Company (GE), the reactor vendor, issued j Rapid Information Communication Services Information Letter (RICSIL) 054, )

                     " Core Support Shroud Crack Indications," on October 3,1990, to all owners of 4

GE BWRs. The RICSIL sunniarized the cracking found in the overseas reactor and recommended that at the next refueling outage-plants with high-carbon-type 304

stainless steel shrouds perform a visual examination of the accessible areas of the seam welds and associated heat-affected zone (HAZ) on the inside and i

outside surfaces of the shroud. l l

Subsequently, a number of domestic BWR licensees performed visual examinations of their core shrouds in accordance with the recommendations in GE RICSIL 054 i or in GE Services Information Letter (SIL) 572, which was issued in late 1993 i to_ incorporate domestic inspection experience. Of the inspections performed l to date, significant cracking was reported at several plants. The combined industry experience from these plants indicates that both axial and j circumferential cracking can occur in the core shrouds of GE designed BWRs.

1 l On July 25, 1994 the NRC issued Generic Letter (GL) 94-03, "Intergranular j Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors," to all BWR licersees (with the exception of Big Rock Point, which does not have a i core shroud) to address the potential for cracking in their core shrouds. GL 94-03 requested BWR licensees to take the following actions with respect to

]                    their corsi shrouds:

inspect their core shronds no later than the next scheduled refueling i outage; 1 l 1 l

{'

  ..g perform a safety analysis supporting continued operation of the facility until the inspections are conducted; develop an inspection plan which addresses inspections of all shroud welds, and which delineates the examination methods to be used for the inspections of the shroud, taking into consideration the best industry technology and inspection experience to date on the subject; develop plans for evaluation and/or repair of the core shroud; and work closely with the BWROG on coordination of inspections, evaluations, and repair options for all BWR internals susceptible to intergranular stress corrosion cracking.

The PECO Energy Company (PEco), the licensee for the Peach Bottem Atomic Power Station Unit 2 (PBAPS 2), responded to GL 94-03 on August 24, 1994 (Reference 1). Part of the licensee's response included PEco's inspection scope for the planned inspection of the PBAPS 2 core shroud, scheduled for refueling outage (RFO) 2R10, which commenced on September 16, 1994. PEco also submitted an i analysis of its proposed modification for the shroud circumferential welds.

• i This modification was not implemented during the Unit 2 RF0 2R10. The staff's evaluation of PEco's proposed modification will be addressed in a separate Safety Evaluation Report (SER).

l 2.0 EVALUATION OF THE LICENSEE'S RESPONSE TO GL 94-03 PEco scheduled and performed comprehensive inspections of the PBAPS 2 core shroud during the unit's RF0 2R10, which commenced on September 16, 1994. The following gives the staff's assessment of the susceptibility of the PBAPS 2 core shroud, the scope of the inspection completed during RF0 2R10, and the licensee's assessment of identified cracking. 2.1 Susceptibility of the PBAPS 2 Core Shroud to IGSCC The core shroud cracks which are the subject of GL 94-03, result from intergranular stress corrosion cracking (IGSCC) which is most often associated with sensitized material near the component welds. IGSCC is a time-dependent phenomena requiring a susceptible material, a corrosive environment, and a tensile stress within the material. Industry experience has shown that austenitic stainless steels with low carbon content are less susceptible to IGSCC than stainless steels with higher carbon content. BWR core shrouds are constructed from either type 304 or 304L stainless steel. Type 304L stainless steel has a lower carbon content that type 304 stainless steel. During the shroud fabrication process when the sections of the core shroud are welded together, the heating of the material adjacent to the weld metal sensitizes the material. Sensitization involves carbon diffusion out of solution forming carbides at grain boundaries upon i moderate heating. The formation of carbides at the grain boundaries depletes l

1 i l

-3_

t i i the chromium in the adjacent material. Since the corrosion resistance of } stainless steel is provided by the presence of chromium in the material, the area adjacent to the grain boundary depleted of chromium is thereby i susceptible to corrosion. Increased material resistance to IGSCC will result if the carbon content is kept below 0.035%, as specified for type 304L grade 3 material. J

• Currently available inspection data indicate that shrouds fabricated with forged ring segments are more resistant to IGSCC than rings constructed from

! welded plate sections. The current understanding for this difference is j related to the surface condition resulting from the two shroud fabrication

processes. Welded shroud rings are constructed by welding together arcs

machined from rolled plate. This process exposes the short transverse direction in the material to the reactor coolant. Elongated grains and

stringers in the material exposed to the reactor coolant environment are

{ believed to accelerate the initiation of IGSCC. i Water chemistry also plays an important role in regard to IGSCC ] susceptibility. Industry experience has shown that plants which have operated with a history of high reactor coolant conductivity have been more, susceptible to IGSCC than plants which have operated with lower conductivities . Furthermore, industry experience has shown that reactor coolant systems (RCSs) i which have been operated at highly positive, electrochemical potentials (ECPs) j have been morg susceptible to IGSCC than RCSs that have been operated at more 1 negative ECPs . The industry has made a considerable effort to improve water ] chemistry at nuclear facilities over the past 10 years. Industry initiatives have included the introduction of hydrogen water chemistry as a means of lowering ECPs i.e., making the ECPs more negative) in the RCS. The i effectiveness o(f hydrogen water chemistry in reducing the susce)tibility of core shrouds to IGSCC initiation has not been fully evaluated; iowever, its effectiveness in reducing IGSCC in recirculation system piping has been l demonstrated. l Welding processes can introduce high residual stresses in the material at the 1 j ' Conductivity is a measure of the anionic and cationic content of

liquids. As a reference, the conductivity of pure water is ~0.05 s/cm.

1 Reactor coolants with conductivities below 0.20 us/cm are considered to be ! relatively ion free; reactor coolants with conductivities above 0.30 s/cm are l considered to have a relatively high ion content. 2 4 The electrochemical potential (ECP) is a measure of a material's j susce)tibility to corrosion. In the absence of an externally applied current, and tierefore, for reactor internals in the RCS, the electrochemical potential ! is equal to the open circuit potential of the material. Industry experience

has shown that crack growth rates in reactor internals are low when the j ECP s ~-0.230 volts.

i j

}

c..

i o j _4_ i

weld joint. The high stresses result from thermal contraction of the weld metal during cooling. A higher residual tensile weld stress will increase the j material's susceptibility to IGSCC. Although weld. stresses are not easily 4

i quantified, previous investigation into weld stresses indicate that tensile i stresses on the weld surface may be as high as the yield stress of the material. The stress decreases to compressive levels in the center of the ] welded section. j PECo has reviewed the materials, fabrication and operational histories of the ) PBAPS ? core shroud and has submitted this information to the staff in their , response to GL 94-03. The PBAPS 2 plant-specific susceptibility factors are ) sumarized below: i i i . i The shroud support, top guide support, and core support plate rings are fabricated from two welded 304 stainless steel, forged ring segments, i 1 with carbon contents of ~0.030%. The shroud shell region was fabricated l by welding rolled 304 stainless steel plates together. The carbon i content of the PBAPS 2 shroud plates are in the range of 0.050 - 0.065%.  ! l - Welding of the shroud plates and rin s for circumferential welds H1'- H6 was accomplished by submerged arc we ding using ER308 filler metal, a i Welding of the bi-metallic weld, H7, was accomplished by gas metal arc welding using filler metal 82. Weld residual stress levels resulting j from these fabrication processes are high. PBAPS 2 operated at high reactor coolant ionic content levels dLring the

~

initial years of operation. The initial five year average coolant conductivity for PBAPS 2 was 0.593 pS/cm, which is considerably higher j than the average for other U.S. BWRs (where the conductivities range from ~0.123 pS/cm to 0.717 pS/cm, and average - 0.340 yS/cm), a PBAPS 2 has operated for 11.8 cumulative years at full power, which is j slightly above the median for U.S. BWRs (range is 3.7 years - 17.8 j years, with a median of 10.8 years). i The PBAPS 2 and Peach Bottom Unit 3 (PBAPS 3) reactors have operated for I approximately the same amount of time at full power, and have in common a i history of operation with high ionic content reactor coolants during the i initial five years of power operation. As a basis for comparison, previous

inspections of circumferential and vertical welds in the PBAPS 3 core shroud revealed the existence of a moderately sized crack (-105 inches in length) i i

along the lower heat affected zone of the shroud's H3 circumferential weld, in 1 1 addition to some less significant cracking at the H1 and H4 weld locations. I From the perspective of materials and fabrication methods, the PBAPS 2 core shroud was fabricated in the same manner as was the PBAPS 3 core shroud. The

 !                 Boiling Water Reactor Vessels & Internals Project (BWRVIP) has classified the

PBAPS 2 core shroud as a susceptible Category "C" shroud. The staff finds

;.                 that the BWRVIP's categorization of the PBAPS 2 core shroud is acceptable

l i j .. l . i and considers the core shroud at PBAPS 2 to be as susceptible to IGSCC as the

;              core shroud in the PBAPS 3 sister unit.

t 4 2.2 Insnection of the Peach Bottom Unit 2 Core Shroud ( By letter dated November 7, 1994, PEco submitted the PBAPS 2 core shroud ) inspection scope, examination results and their flaw evaluation. i

!              2.2.1 Scone of Core Shroud Inspection j

The P8APS 2 shroud examinations were performed using the ultrasonic testing (UT) methods developed by the General Electric Corporation (GE). The UT examinations utilized GE's Smart-2000 Data Acquisition System and the GE 00 i Tracker and suction cup scanners. The extent of the )lanned UT examinations ! included all accessible portions of circumferential siroud welds HI - H7. The l UT examinations were performed using three UT transducers, a 45* shear wave transducer, a 60' longitudinal wave transducer, and a creeping wave transducer i which was used to pick up surface indications. The cree)ing wave transducer l i was not used on the H3 weld due to equipment failure. Tie licensee also l performed some additional enhanced VT-1 examinations of shroud weld H6, whicN. 1 was highly obstructed by the proximity of the jet pumps and therefore highly.' < , inaccessible to the GE UT equipment. The licensee indicated that it had - j completed the following PBAPS 2 core shroud UT examinations:  ; l 33% of the length (230") of weld H1, distributed over 66% of the weld's 5 circumference, j - 84% of the length (583") of weld H2, t 88% of the length (574") of weld H3, 89% of the length (580") of weld H4, I - 83% of the length (540") of weld H5, j - 10% of the length (148") of weld H6, plus an additional 13% of weld H6 by enhanced VT-1 examination techniques, and '1 - 9% of the length (59") weld H7, in areas which were accessible by way of the access hole covers. { 2.2.2 Core Shroud Examination Results j 4 The following summarizes the cracking identified at each weld during the examination-of the PBAPS 2 core shroud. H1 Wald - The examination detected 11 indications by UT using 45'S/60'RL i transducers, totalling 33.93 inches, with a maximum length of 4.75 inches and a maximum depth of 0.74 inches at Indication #7; 4 - H2 Weld - Examinations were negative for indications; H3 Weld - 19 indications were detected by UT using 45'S/60'RL transducers, totalling 68.48 inches, with the maximum length being 8.75

!                      inches at Indication #16 (indications were not depth sized).

4 1

                                                                                      ~            -

1 l ll

     ,                                                      6-I                                                                                                     l l

i - H4 Weld - 8 indications were identified, totalling 11.46 inches, with ' the maximum length of 5.76 inches at Indication #4 (indications were not depth sized) as detected by UT using 45'S/60'RL transducers, and remaining seven indications detected by UT creeping wave measurements; l l H5 Weld - 1 indication 2.28 inches in length was detected by UT creeping i 1 wave (indication was not depth sized); H6 Weld - 1 indication was detected by UT using 45'S/60'RL transducers, 4.73 inches in length and 0.45 inches in depth; H7 Weld - examinat< ons were negative for indications. i The licensee's inspections of welds H6, the core support ring-to-lower shroud j weld, H7, the lower shroud-to-shroud support cylinder weld, and H8, the shroud l 1 support cylinder-to-jet pump support ledge weld, were conducted through accessible areas of the access hole covers. Interference from jet pump assemblies, the reactor core, and other internals located at lower vessel elevations limited access to the lower shroud welds. The licensee's inspection plan is consistent with the staff's inspection of all accessible shroud weld areas. position recommending a 100%  ; i 2.2.3 Assessment of the PBAPS 2 Core Shroud Insoection Results . Flaws identified in welds receiving a comprehensive examination during the fall 1994 RF0 were evaluated in accordance with the methodology outlined in the "BWR Core Shroud Inspection and Flaw Evaluation Guidelines" (Reference 2). These guidelines closely follow the flaw evaluation guidelines found in Section XI of the ASME Code. The staff has reviewed the BWRVIP evaluation guidelines and approves of the use of the quantitative assessment methods. The licensee's evaluations were based on the following assumptions and conditions:

                       . For welds that were largely accessible to examinations and for which comprehensive examinations were performed, all as-found indications were assumed to be through-wall, which removed the necessity for depth characterization. Additionally, any inaccessible areas were assumed to contain through-wall indications over their entire inaccessible lengths.
                       . For welds that were predominantly inaccessible to examination, conditions found within the inspected regions were extrapolated over the entire weld areas that were inaccessible to examination equipment. The extrapolated conditions were then evaluated for structural integrity.

Thus, evaluations of the H1 and H6 welds, in which indications were found and which were sized for depth, were based upon the assumption that the majority of the welds' circumferences contained indications.

                       . For the H7 weld, in which no indications were found, calculations were performed to calculate the depth which could be tolerated assuming a 360* crack existed in the weld.

4 i

         .-                                                            _7-i i

As-found crack lengths were adjusted for crack growth, non-destructive examination uncertainties, and crack proximity factors in accordance with the guidelines (Reference 2). l Inspection results for those welds receiving comprehensive inspections were j compared to the initial screening criteria established in GENE 523-176-1293,

                               " Evaluation and Screening Criteria for the Peach Bottom Unit 2 Shroud" 1                               (Reference 3), and 17 unacceptable, evaluated for safety marg' ins using limit

load methodology found in the "BWR Core Shroud Inspection and Flaw Evaluation i Guidelines" (Reference 2). The inspection results of the H3 and H4 welds were also subject to evaluation using linear elastic fracture machanics methods to l

account for high neutron fluences which are comon at these weld elevations.

l Safety margins were calculated against the most limiting design basis loading l conditions, derived in GENE 523-176-1293, " Evaluation and Screening Criteria for the Peach Bottom Unit 2 Shroud" (Reference 3). This equated to use of faulted condition loadings for evaluations of circumferential welds HI - H5, and upset condition loadings for circumferential welds H6 and H7. For all postulated loadings the licensee showed that the loadings conditions for the as-found conditions in welds H1 - H7 were less than the ASME Code stress intensity allowables. The licensee's evaluations of the P8APS 2 core shroud indicate that the shroud will maintain its structural integrity even under the most severe loading conditions for a given shroud ~ weld location. The staff has reviewed the licensee's methodology, and has determined that the licensee's method of evaluating the PBAPS 2 core shroud is acceptable and that the licensee's evaluation results justify operation of the PBAPS 2 unit for the next operating cycle.

3.0 CONCLUSION

S Based on a review of the PBAPS 2 core shroud materials, fabrication )rocesses and operating history the staff concludes that the licensee's core siroud is susceptible to IGSCC. PECo completed an examination of the PBAPS 2 core shroud during RF0 2R10. The licensee's evaluation of the PBAPS 2 core shroud indicates that the P8APS 2 core shroud will maintain sufficient structural margins to justify operation of the PBAPS 2 reactor for another operating cycle without necessitating a modification of the PBAPS 2 core shroud. 4.0 OUTSTANDING ISSUES / FUTURE ACTIONS The licensee's difficulty inspecting the some of the circumferential core shroud welds is not unique to this plant. It should be noted that the industry is currently encountering difficulties performing comprehensive inspections of lower shroud weldt, due to NDE equipment accessibility problems. The staff urges licensees to work with the members of the EPRI NDE Center in

I

  ..} .                                                                                                                                                     1 order to develop improved tooling for inspections of lower shroud welds and/or        '

lower vessel regions which are highly obstructed. Should improved inspections l techniques become available, the staff recomendation is for licensee's to l reinspect the lower shroud welds at the earliest opportunity. l i

5.0 REFERENCES

1. Letter from G. A. Hunger, Jr., PECO to the U.S. Nuclear' Regulatory Commission forwarding the " Peach Bottom Atomic Power Station, Units 2 and 3, Limerick Generating Station Units 1 and 2 Response to Generic Letter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shroud in Boiling Water Reactors," dated August 24, 1994.

2. Letter From C. D. Terry, Executive Chairman, Assessment Committee, BWR Vessel & Internals Project, to the U.S. Nuclear Regulatory Commission forwarding the "BWR Core Shroud Inspection and Evaluation Guidelines," dated September 2, 1994.

3. M. L. Herrera and S. Raganath, " Evaluation and Screening Criteria for the Peach Bottom Unit-2 Shroud," Rev. O, (GENE-523-276-1093) dated December 13, 1993.

l Principal Contributor: J. Medoff Date: February 6, 1995

j I  : gn arc [. Jq UNITED STATES E j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

      \ . . . . . 8,o

• January 29, 1996 Mr. George A. Hunger, Jr.

Director-Licensing, MC 62A-1 PECO Energy Company , Nuclear Group Headquarters l Correspondence Control Desk P.O. Box No. 195 Wayne, PA 19087-0195

SUBJECT:

GENERIC LETTER (GL) 94-03, "INTERGRANULAR STRESS CORR 0SION CRACKING 0F CORE SHROUDS IN BOILING WATER REACTORS," PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 3 (TAC NO. M90106)

Dear Mr. Hunger:

By letter dated August 24, 1994, the PEC0 Energy Company (PECO) provided its response to Generic Letter (GL) 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors," for the Peach Bottom Atomic Power Station, Units 2 and 3. The NRC staff requested in GL 94-03 that licensee's take the following actions with respect to their core shrouds: (1) inspect their core shrouds in their boiling water reactor (BWR) plants no later than the next refueling outage; (2) perform materials-related and plant-specific consequence safety analyses with respect to their core shrouds; (3) develop core shroud inspection plans which address inspection of all core shroud welds and which takes into account the latest available inspection technology; (4) develop plans for evaluation and/or repair of their core shrouds; and (5) work closely with the BWR Owners Group (BWROG) with respect to addressing intergranular stress corrosion cracking of BWR internals. The NRC staff requested that licensee's submit, under oath or affirmation, the following information in response to GL 94-03 within 30 days of the date of issuance: (1) a schedule for inspection of their core shrouds; (2) a safety analysis, including a plant-specific safety analysis as appropriate, which supports continued operation of the facility until inspections are conducted; (3) a drawing (s) of the core shroud configurations; and (4) a history of shroud inspections completed to date. The NRC staff also requested that licensee's submit, under oath or affirmation, no later than 3 months prior to performing their core shroud inspections, their scope for inspection of their core shrouds and their plans for evaluating and/or repairing their core shrouds based on their inspection results. The NRC staff further requested licensee's to submit, under oath or affirmation, their core shroud inspection results and flaw evaluation within 30 days of completing their shroud examinations.

                           ,      Y

' *t G. Hunger, Jr. PEC0 recently completed the Peach Bottom Unit 3 (PBAPS 3) core shroud examinations, which were performed during the recently completed refueling outage 3R010, and which were performed per the actions requested by GL 94-03. The staff has received the November 3,1995 submittal containing the results and evaluation of the PBAPS 3 core shroud examinations which were performed during refueling outage 3R010. Your evaluation of the PBAPS 3 core shroud indicates that the PBAPS 3 core shroud will maintain sufficient structural margins to justify operation of the PBAPS 3 reactor for another operating i cycle without necessitating a modification of the PBAPS 3 core shroud. The l staff has reviewed PECO's ins)ection and evaluation methodology and finds them l acceptable for operation of tie Unit 3 core shroud for the duration of the current operating cycle. The staff's evaluation of PEC0's November 9,1995 shroud inspection and analysis submittal is enclosed. Staff action for PECO's response to GL 94-03 for Unit 3 is completed and TAC M90106 is closed. Sincerely,

                                                                               /S/

l Joseph W. Shea, Project Manager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-278 l 1

Enclosure:

Safety Evaluation cc w/ encl: See next page DISTRIBUTION: PUBLIC JStolz JShea WPasciak, RGN-I Docket File JMedoff CECarpenter JZwolinski PDI-2/ Reading JStrosnider M0'Brien SVarga 0GC ACRS

• Previous Concurrence CFFICE uffI! kl.2/fyt EMCS/BC
• Cl*2/0 NAME MO rien J JStrosnider (2 DATE f /% \ /tC(% 01/18/M / l /d%

UFFICIAL RECORD COPY l FILENAME: Ga \$ HEN \U3SHROUO. INS

's i _ G. Hunger, Jr. PECO recently completed the Peach Bottom Unit 3 (PBAPS 3) core shroud examinations, which were performed during the recently completed refueling outage 3R010, and which were performed per the actions requested by GL 94-03. The staff has received the November 3,1995 submittal containing the results and evaluation of the PBAPS 3 core shroud examinations which were performed during refueling outage 3R010. Your evaluation of the PBAPS 3 core shroud indicates that the PBAPS 3 core shroud will maintain sufficient structural . margins to justify operation of the PBAPS 3 reactor for another operating cycle without necessitating a modification of the PBAPS 3 core shroud. The staff has reviewed PEC0's inspection and evaluation methodology and finds them acceptable for operation of the Unit 3 core shroud for the duration of the current operating cycle. The staff's evaluation of PEC0's November 9,1995 shroud inspection and analysis submittal is enclosed. Staff action for PEC0's response to GL 94-03 for Unit 3 is completed and TAC M90106 is closed. Sinc ely,

                                             )

Jo seq W. Shea, Project Manager Pr>jact Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation-Do~cket No. 50-278

Enclosure:

Safety Evaluation cc w/ encl: See next page

                                                                                     )
                                                                                     )

Mr. George A. Hunger, Jr. Peach Bottom Atomic Power Station, j PECO Energy Company Units 2 and 3 i cc: l l J. W. Durham, Sr., Esquire Mr. Rich R. Janati, Chief Sr. V.P. & General Counsel Division of Nuclear Safety PECO Energy Company Pennsylvania Department of 2301 Market Street, S26-1 Environmental Resources Philadelphia, Pennsylvania 19101 P. O. Box 8469 Harrisburg, Pennsylvania 17105-8469 PECO Energy Company ATTN: Mr. G. R. Rainey, Vice President Board of Supervisors l Peach Bottom Atomic Power Station Peach Bottom Township Route 1, Box 208 R. D. #1 Delta, Pennsylvania 17314 Delta, Pennsylvania 17314 PECO Energy Company Public Service Comission of Maryland ATTN: Regulatory Engineer, A4-SS Engineering Division Peach Bottom Atomic Power Station Chief Engineer Route 1. Box 208 6 St. Paul Centre Delta, Pennsylvania 17314 Baltimore, MD 21202-6806 Resident Inspector Mr. Richard McLean U.S. Nuclear Regulatory Comission Power Plant and Environmental Peach Bottom Atomic Power Station Review Division P.O. Box 399 Department of Natural Resources l Delta, Pennsylvania 17314 B-3, Tawes State Office Building Annapolis, Maryland 21401 Regional Administrator, Region I U.S. Nuclear Regulatory Comission Dr. Judith Johnsrud 475 Allendale Road National Energy Comittee King of Prussia, Pennsylvania 19406 Sierra Club 433 Orlando Avenue Mr. Roland Fletcher State College, PA 16803 Department of Environment 201 West Preston Street Robin Dyle, Technical Chairman Baltimore, Maryland 21201 BWRVIP Assessment Task l Southern Nuclear Operating Company Warren Bilanin, EPRI Task Manager Post Office Box 236 3412 Hillview Ave. 40 Inverness Center Parkway Palo Alto, CA 94303 Birmingham, AL 35201 I A. F. Kirby, III External Operations - Nuclear i Delmarva Power & Light Company l P.O. Box 231 l Wilmington, DE 19899

__ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ . _ . . . _ _ ~_ _ _ _ __ .._ .. _ _ .. _ _ _ l . po Cto y*  % UNITED STATES ' j j NUCLEAR REGULATORY COMMISSION

   $             $                                 WASHINGTON, D.C. 20665 0001                                                  .

N ...../ l l SAFETY EVALUATION EY THE OFFICE OF NUCLEAR REACTOR REGULATION l l RESPONSE TO GENERIC LETTER 94-03 l PEACH BOTTOM ATOMIC POWER STATION. UNIT 3 PECO ENERGY COMPANY DOCKET NO. 50-278

1.0 BACKGROUND

1 The core shroud in a Boiling Water Reactor (BWR) is a stainless steel cylindrical component within the reactor pressure vessel (RPV) that surrounds I the reactor core. The core shroud serves as a partition between feedwater in l the reactor vessel's downcomer annulus region and the cooling water flowing up  : through the reactor core. In addition, the core shroud provides a refloodable volume for safe shutdown cooling and laterally supports the fuel assemblies to i maintain control rod insertion geometry during operational transients and accidents. In 1990, crack indications were observed at core shroud welds located in the beltline region of an overseas BWR. This reactor had completed approximately 190 months of power operation before discovery of the cracks. As a result of this discovery, General Electric Company (GE), the reactor vendor, issued Rapid Information Communication Services Information Letter (RICSIL) 054,

        " Core Support Shroud Crack Indications," on October 3,1990, to all owners of GE BWRs. The RICSIL summarized the cracking found in the overseas reactor and recommended that at the next refueling outage plants with high-carbon-type 304 stainless steel shrouds perform a visual examination of the accessible areas of the seam welds and associated heat-affected zone (HAZ) on the inside and outside surfaces of the shroud.

Subsequently, a number of domestic BWR licensees performed visual examinations of their core shrouds in accordance with the recommendations in GE RICSIL 054 or in GE Services Information Letter (SIL) 572, which was issued in late 1993 to incorporate domestic inspection experience. Of the inspections performed to date, significant cracking was reported at several plants. The combined industry experience from these plants indicates ~ that both axial and circumferential cracking can occur in the core shrouds of GE designed BWRs. , On July 25, 1994 the NRC issued Generic Letter (GL) 94-03, "Intergranular

Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors," to all Enclosure l

4 BWR licensees (with the exception of Big Rock Point, which does not have a core shroud) to address the potential for cracking in their core shrouds. GL 94-03 requested BWR licensees take the following actions with respect to their core shrouds: inspect their core shrcuds no later than the next scheduled refueling outage; perform a safety analysis supporting continued operation of the facility until the inspections are conducted; l develop an inspection plan which addresses inspections of all shroud l welds, and which delineates the examination methods to be used for the inspections of the shroud, taking into consideration the best industry t~hnology and inspection experience to date on the subject; I develop plans for evaluation and/or repair of the core shroud; and

                            - work closely with the BWROG on coordination of inspections, evaluations, and repair options for all BWR internals susceptible to intergranular                 ,

stress corrosion cracking.  ! l The PECO Energy Company (FECOs, the licensee for the Peach Bottom Atomic Power Station, Unit 3 (PBAPS 3), responded to GL 94-03 on Au3ust 24, 1994 (Reference 1). Part of the licensee's response included PECO's inspection scope for the planned re-inspections of the PBAPS 3 core shror.6, which were scheduled for refueling outage (RFO) 3R010 in the fall of 1955. PECO provided additional l information regarding the 3R010 shroud inspection plans in a letter dated June l 16, 1995. The licensee previously completed an inspection of the PBAPS 3 core shroud during the refueling outage 3R09 during the fail of 1993. The General l Electric Nuclear Energy Division formally submitted the ex&mination results l and assessment of core shroud structural integrity to the NRC by letter dated December 3, 1993. PECO amended the results and assessment by letter dated March 14, 1994. The NRC staff's review of the results of PECO's inspection and assessment is documented in a safety evaluation dated February 6,1995. 2.0 Ey4j,'IATIGN OF THE LICENSEE'S RESPONSE TO GL 94-03 l PECO scheduled and performed comprehensive inspections of the PBAPS 3 core shroud during the unit's RF0 3R010, which commenced in September, 1995. The following gives the staff's assessment of the susceptibility of the PBAPS 3 core shroud, the scope of the inspection completed during RF0 3R010, aM the ! licensee's assessment of identified t . king. i l ! l I ! l

1 l .

t- _ _ _ _ _ - _ -a

i a, i l -3 - 1 j 2.1 Susceptibility of the PBAPS 3 Core Shroud to Intergranular Stress Corrosion Cracking (IGSCC) The core shroud cracks which are the subject of GL 94-03, result from IGSCC ' which is most often associated with sensitized material near the ccmponent welds. IGSCC is a time-dependent phenomena requiring a susceptible material, j a corrosive environment, and a tensile stress within the material. ) Industry experience has shown that austenitic stainlest steels with low carbon i cor. tent are less susceptible to IGSCC than st..nless steels with higher carbon i content. BWR core shrouds are constructed from either type 304 or 304L l stainless steel. Type 304L stainless steel has a lower carbon content than

type 304 stainless steel. During the shroud fabrication process when the sections of the core shreud are welded together, the heating of the material

! adjacent to the weld metal sensitizes the material. Sensitization involves carbon diffusion out of solution forming carbides at grain boundaries upon moderate heating. The formation of carbides at the grain boundaries depletes 8 the chromium in the adjacent material. Since the corrosion resistance of ) stainless steel is provided by the presence of chromium in the material, the 4 area adjacent to the grain boundary depleted of chromium is thereby ! susceptible to corrosion. Increased material resistance to IGSCC will result if the carbon content is kept below 0.030%, as specified for type 304L grade j material. Currently available insemion data indicates that shrouds fabricated with forged ring segments are :m,re resistant to IGSCC than rings constructed from welded plate sections. The current understanding for this difference is related to the surface condition resulting from the two shroud fabrication processes. Welded shroud rings are constructed by welding together c.rcs machined from rolled plate. This process exposes the short transverse direction in the material to the reactor coolant. Elongated grains and stringers in the material exposed to the reactor coolant environment are believed to accelerate the initiation of IGSCC. Water chemistry also plays an important role in regard to IGSCC susceptibility. Industry experience has shown that plants which have operated with a history of high reactor coolant conductivity have been more gusceptible to IGSCC than plants which have operated with lower conductivities. Furthermore, industry experience has shown that reactor coolant systems (RCSs) which have been operated at highly positive, electrochemical potentials (ECPs)

                        ' Conductivity is a measure of the anionic and cationic content of liquids. As a reference, the conductivity of pure water is ~0.05 s/cm.

Reactor coolants with conductivities below 0.20 s/cm are considered to be relatively ion free; reactor coolants with conductivities above 0.30 s/cm are considered to have a relatively high ion content.

have been more, susceptible to IGSCC than RCSs that have been operated at more i negative ECPs. The industry has made a considerable effort to improve water

chemistry at nuclear facilities over the past 10 years. Industry initiatives I j have included the introduction of hydrogen water chemistry as a means of I q lowering ECPs (i.e., making the ECPs more negative) in the RCS. The l

effectiveness of hydrogen water chemistry in reducing the susceptibility of 1 core shrouds to IGSCC initiation has not been fully evaluated; however, its .

! effective- ' in reducing IGSCC in recirculation system piping has been j demonstrata. ! Welding processes can introduce high residual stresses in the material at the { weld joint. The high stresses result from thermal contraction of the weld l metal during cooling. A higher residual tensile weld stress will increase the

material's susceptibility to IGSCC. Although weld stresses are not easily 4

quantified, previous investigation into weld stresses indicate that tensile ! stresses on the weld surface may be as high as the yield stress of the ! material. The stress decreases to compressive levels in the center of the i welded section. i PECO has reviewe - /6- s_s5M l icTa'rd H.L hards, NDE Technician Date i obile NDE Laboratory l Division of Reactor Safety l

                                                                         ~
                                                                                    /c/2S/* 94-Patrick M. Peterson, NDE Technician      Date Mobile NDE Laboratory Division of Reactor Safety APPROVED:                                                            /M28/9Y E. Harold Gray, Chief                     Date Mobile NDE Laboratory Division of Reactor Safety Insoection Summary: An announced inspection was conducted at Peach Bottom Atomic Station (PBAS), September 26 through October 7,1994, using the Nuclear Regulatory Commission (NRC) Nondestructive Examination (NDE) Mobile Laboratory (ML) (Report No. 50-277/50-278/94-404). The purpose of the NDE Mobile Laboratory is to perform independent nondestructive examinations and evaluations of components, systems and weldments to assure that examinations performed are in compliance with codes, standards and regulatory requirements.

Areas Insoected: The inspectun included performance of nondestructive examinations on safety-related piping welds, pipe hangers and supports selected from the residual heat removal (RHR), essential service water (ESW), reactor water clean-up (RWCU), feedwater (FW) and recirculation (RCS) systems. Areas examined during this inspection included a review of the licensee's in-service inspection (ISI) program application. Also included in this

                '              1 941 ADOCK 050002 7

__ _ _ _ _ - - - _ _ - - ~ _ _ _ _ . _ -_ .- _ __ i inspection, was a review of the flow accelerated corrosion (FAC) program for high energy piping, the corrosion control program for the Emergency Service i Water (ESW) system and the vessel core shroud visual (VT) and ultrasonic (UT) ] l examinations. Results: Within the expected normal variations in examination techniques, the results of the NDE evaluations performed by the NRC essentially agreed with the results obtained by the licensee. Three items of weakness were identified in the ISI/NDE program application. These were not marking the weld centerline on welds for UT as part of the ISI program, not finding or recording a geometric reflector in excess of 50% of DAC while conducting UT per the ASME Code on a RWCU system weld, and hav.ing radiographs that show i signs of aging in storage for work performed after original construction. 1 l l l l I ii

DETAILS 1.0 IMEPENDENT MEASUREMENTS - NRC N0NDESTRUCTIVE EXAMINATION AND QUALITY l RECORDS REVIEW 0F SAFETY-RELATED SYSTEMS (73753) l During the period of September 26 through October 7, 1994, an onsite independent inspection was conducted at Peach Bottom Atomic Station, Units 2 and 3. The inspection was conducted by NRC inspectors and contractors. The i objective of this inspection was to assess the adequacy of the licensee's l inservice inspection (ISI) program application, implementation of the flow ! accelerated corrosion (FAC) program, measurements to determine the status of l l corrosion in the ESW system and the condition of radiographs held in storage 1 of plant piping welds. A part of the inspection was accomplished by l l duplicating a sample of those examinations performed by the licensee as required by regulations and codes, evaluating the results. The inspection included review of the ISI program and its implementation, the FAC program and NDE procedures used to implement these programs. Section 4.0 of this report  ! contains a listing of the specific welds inspected. l The Code of Federal Regulations (CFR), Title 10, Part 50.55a (10 CFR 50.55a(g)), requires ISI of safety-related equipment to identify system j degradation. Before the licensee generated program of inspection is applied ) to the equipment, it must be submitted for review and approval by the NRC under the authority embodied in 10 CFR 50.55a(g)(4)(iv). The required inspections are detailed in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, for Inservice Inspection per 10 CFR 50.55a(b). The NRC inspection described in this report was made using the NDE Mobile Laboratory. The NDE Mobile Laboratory is capable of l independently performing the examinations required of the licensee. This capability provides the NRC with unique insights into the licensee's inservice inspection program and on a sampling basis, the adequacy and accuracy of the licensee's specific NDE inspections. In summary, the scope of this inspection was to review the portions of the ISI/NDE program, assess its implementation and to perform NDE on portions of the plant systems. Results: The inspection team concluded the ISI program at Peach Bottom is well planned, well controlled, and well executed, meeting the minimum requirements of the ASME Code. However, during the inspection, three weaknesses were noted. A weakness is a condition that without attention could develop into a safety problem or regulatory issue. Weaknesses are identified for licensee review and corrective action as appropriate. The weaknesses are described in the specific sections of this report. 1.1 Nondestructive Examination (NDE) Insoection Hanaer/Suonort (57050) l The visual inspection of pipe hanger / support 10 'GB-H154 was done per NRC Procedure NDE-10, Rev. O, Appendix A and B, in conjunction with Peach Bottom Site Procedure MAG-CG-407, Revision 2, quality control documents, and associated isometric / drawings. The accessible surface area and adjacent base metal for a distance of one-half inch on either side of the weld was examined.

2

       ' Component integrity was also examined, including proper installation, configuration or modification of supports, evidence of mechanical or structural damage, corrosion, bent, missing or broken members. During the visual examination af pipe hanger supports, the as-found settings on the spring hanger 10GB H154 were found to be out of tolerance and debris was observed inside both spring cans. These unacceptable conditions were previously noted and documented by the licensee's inspector on Examination Summary Sheet No. 202950. An action request was written for this item and processed to correct the unr.cceptable condition. The inspections by the NRC closely matched those of the licensee.

Results:' One unacceptable condition was idertified by the licensee and confirmed by the NRC inspector. The condition is not a concern in that it was identified, properly documented and being processed for evaluation and corrective action. Visual Examination (57050) Fifteen (15) safety-related pipe weldments and adjacent base material (1/2 i inch on either side of the weld) were visually examined in accordance with NRC Procedure NDE-10, Rev. O, Appendix A, Peach Bottom Site Procedure MAG-CG-401, Revision 2, quality control documents, isometrics and as-built drawings. Examined during this inspection were ASME Class 1 and 2, pipe weldments selected from the FW, RHR, and RWCU systems. Inspections were performed specifically to identify any cracks or linear indications, gouges, leakage, arc strikes with craters, or corrosion, which may infringe upon the minimum pipe wall thickness and modifications to piping or components. Mirrors, flashlights, and weld gauges were used to aid in the inspection and evaluation. Table 1 lists the specific welds examined. Results: The welds examined were ground for in-service inspection prior to surface and volumetric examinations. The welding and overall workmanship inspected was acceptable. The inspection reports of the licensee reflected the as-found conditions. The licensees examinations matched closely to those of the NRC. No deviations were identified. , Liouid Penetrant Examination (57060) Ten (10) safety-related pipe weldments and adjacent base material (1/2 inch on either side of the weld) were examined using the visible dye, solvent removable method per NRC Procedure NDE-9, Rev. 1, in conjunction with Peach Bottom Site Procedure LP-PE-001, Revision 4. Included in this inspection were ASME Class 1 and 2 stainless and carbon steel pipe weldments selected from the RWCU, FW, and RHR systems. Table 1 lists the specific welds examined. Results: The surface areas examined were adequately prepared for the examination. The licensee recorded the same relevant indications noted by the NRC. The licensees examinations matched closely to those of the NRC. No rejectable indications were identified; no deviations noted.

                                                                    - . - - - - . - - - . - . - ~

4 3 l Maanetic Particle Examination (57070) Five (5) safety-related pipe weldments were examined with the magnetic particle method using NRC Procedure NDE-6, Rev.~1, Peach Bottom Site Procedure l MT-PE-001, Revision 2, associated quality control documents and isometric 1 l drawings. Included in this examination were ASME Class 2 pipe weldsents from l I the FW system. Table 1 lists the specific welds examined. Results: The surface areas examined were adequately prepared for the

examinations. There were no recordable indications found by the NRC. The l licensees examinations matched closely to those of the NRC. l Ultrasonic Examination (Manual) (57080)

Twelve (12) safety-related pipe weldments were ultrasonically examined using NRC Procedure NDE-1, Rev. 1, in conjunction with Peach Bottom Site Procedure UT-PE-002, Rev. 8, and associated isometric drawings and ultrasonic inspection reports. Included in this examination were ASME Class 1 and 2 pipe weldments selected from the RHR, RWCU, and FW systems. To obtain the greatest possible repeatability in performing the NRC independent evaluations, the examinations were performed utilizing ultrasonic units, transducers and cables that l matched, as closely as possible, those used by the licensee. Distance amplitude correction (DAC) curves were established utilizing Peach Bottom calibration standards. Table 1 lists the specific welds examined. Results: One weakness was noted during the ultrasonic inspections. The licensee's contractor failed to find and document a recordable indication greater than 50% of DAC on weld 12-14-5 in the RWCU system. The inspection performed by the licensee meet the ASME Code and the UT Procedure UT-PE-002, Rev. 8. The transducer and wedge used by the licensee differed from that'of the NRC. The NRC's transducer required 6dB less than the licensees to obtain the same response. Also, the exit point was .05" less and the refracted angle was 3* greater (46*) than that of the licensees. The greater angle and shorter exit point provided a greater response from the geometric reflector for the NRC's examination. The weakness in this examination is the selection of equipment and verifying adequate examination coverage. Although the transducer was within tolerance of the Code, multiple variables on the low end of the tolerances made for a less sensitive examination. The other UT examinations matched closely to those of the NRC. No other items of concern were noted. Ultrasonic Examination (Automated) (57080) l I During this inspection period, the NRC Synthetic Aperture Focusing Technique (SAFT) UT System was used to examine a safe-end to nozzle weld (2-BKA-8) on the recirculation system. The SAFT-UT is an automated, computerized ultrasonic system that was developed to demonstr. ate an advanced imaging 1 technology for pipe weld and reactor vessel in-service inspection. The SAFT , system presents a visual image of ultrasonic reflectors that may be located in l piping or pressure vessels. l l 1 l

l l 4 Synthetic Aperture Focusing refers to a process in which the focal properties of a large-aperture focused transducer are synthetically generated from data collected over a large area using a small transducer with a divergent sound field. The SAFT-UT system consists of four functional subsystems, the data acquisition, host computer, the real-time processor and the graphics processor that provides the image display to the UT technician for interpretation. Approximately 24 inches of the stainless steel side of the safe-end to reactor vessel nozzle was examined and analyzed using the SAFT-UT system. Results: The information obtained by the SAFT-UT system was compared with computer-based UT data collected by licensee's contractor and found to be closely matched. The only variation was in the image presentation and that is due to the way different automated systems display image presentations. No deviations or recordable indications were identified. Core Shroud l l Generic' Letter 94-03 requires boiling water reactor (BWR) owners to inspect l the core shroud welds for intergranular stress corrosion cracking (IGSCC). PBAS, under the guidance of the BWR Vessel Inspection Project (BWRVIP),  ; performed ultrasonic inspection on the accessible portions of the horizontal j welds. The UT technique used for the inspection was demonstrated to the PECO 1 Energy UT Level III at the Electric Power Research Institute (EPRI), i ! Charlotte, NC. The inspection technique utilized a ~45' shear wave, a 60* j refracted longitudinal wave, and 80* OD creeping wave, in accordance with Procedure UT-PBM-503VI, Rev. O. The examinations were performed from the l  ! outer shroud surface from one or both sides of the horizontal welds where accessible. Enhanced visual examination was performed on portions of the weld H-7 not accessible with UT. The coverage that was afforded to each weld is given below: l H1: 72* H2: 340* H3: 340' H4: 340* H5: 340* H6: 30* H7: 30* The licensee presented the following preliminary results as the percent of the tested weld length that was flawed for the Unit 2 shroud inspections: i H1: 22.5% flawed H2: No flaws detected H3: 12.2% flawed H4: 2% flawed H5: 0.43% flawed H6: 8.5% flawed ( All as 1 indication: 4" long, 0.45" in the forged ring) H7: No indications Results: No items of concern were identified for the UT and VT examination methods applied or the controls in place over the performance of the j examinations. The weld examination volume was noted to be extensive for the weld areas examined. The OD creeping wave transducer was demonstrated to the [ l L___ _ _ _ ,

. \ 5 1 NRC UT technician and to the NRC UT Level III. The evaluation of the results for the shroud is based on the Limit Load for welds H-1, H-2, H-5, H-6, and H-7. Limit Load and Linear Elastic Fracture Mechanics is used for the evaluation for welds H-3 and H-4. The UT data was evaluated by the licensees contractor. The extent of IGSCC cracking was significantly less for the Unit 2 shroud than that previously found in the Unit 3 shroud. This finding is consistent with the relative magnitude of IGSCC found in the past between the two plants. Emeraency Service Water System (ESW) l The ESW system provides cooling water to several safety-related items including the emergency diesel generators (EDG), room and equipment coolers. PECO initiated action request (AR) A0004883 in November 1990 to monitor corrosion damage of ESW piping per the NRC Generic Letter (GL) 89-13. This resulted in the selection of 10 locations in the ESW system for inspection by measurement. In November 1993, a leak was identified in a piping elbow weld near valve 0503A that was not included in the original sample of 10 areas. An additional five areas were then measured by the ' licensee in the areas of valves 0503A and 0503B, and a repair plan was initiated for the leak near valve 0503A involving replacement of a portion of the piping. The NDE ML technicians measured portions of the piping near valves 0503A/B and ESW piping supplying three of four of the plant's emergency diesel generators. The thickness measurements in the vicinity of valves 0503A/B were similar to those obtained by the licensee, and those in the EDG rooms were at or above the required minimum wall thickness of 0.225." Results: No items of concern were identified. Control of Qualification and Certification of NDE Technicians. The NDE Support Group reviews the qualification package for each NDE technician per procedure MAG-CG-440, Rev. O, and signs the data package indicating acceptance to the contractor. The NDE technician qualification packages are independently reviewed by the ASME Code Authorized Nuclear Inservice Inspector (ANII). The NRC inspector reviewed a sample of the NDE technician's data packages and the ANII log to confirm that qualification / certification documentation for NDE technicians performing examinations for the inservice inspection program are being properly reviewed. Results: The process of review and control of the qualifications and certifications for NDE technicians was found to be well controlled. Condition of Stored Radioaraohs A sample of radiographs was examined to determine if they were or were not aging in storage. This examination found a problem of aging for radiographs, by two radiographic contractors in the time period of May 1976 (post-construction) until June of 1980. While the radiographs were still readable, this film aging is considered to be a weakness in the PBAS NDE program. The l l 1 I 1 I

l~~ 1 l l 6 I concern was verified by the licensee by initially viewing the same film sample  ; although the bounds of the problem were not full,y identified as of the i conclusion of this inspection. Flow-Accelerated Corrosion (49001) l Concerns regarding erosion / corrosion (E/C), currently known as flow-l accelerated corrosion (FAC), in balance of plant piping systems has heightened as a result of the December 9, 1986, feedwater piping line rupture that occurred at Surry. This event was the subject of NRC Information Notice 86-106, issued December 16, 1986, and its supplement issued on f February 13, 1987. The licensee's actions with regard to the detection of FAC l in plant components were reviewed with respect to NUREG-1344, " Erosion / l Corrosion Induced Pipe Wall Thinning in U. S. Nuclear Power Plants," dated i ! April 1989, Generic Letter 89-08 issued May 2,1989, and NUMARC Technical l Subcommittee Working Group on Piping and Erosion / Corrosion Summary Report, dated June 11, 1987. Peach Bottom Site Procedure MAG-CG-406, Revision 2, provides the FAC program, plan and inspection summary. This is applicable to components in the feedwater, extraction steam, cross under piping, heater and , main steam reheater drains. The basis for the Peach Bottom FAC program is the  ! guidance of EPRI Report NP-3944, related EPRI reports, and industry input. During this Unit 2 outage,139 FAC inspections were scheduled by the licensee l including 50 large bore pipe, 68 small bore pipe, and 21 baseline examinations. The components were selected for examination using the EPRI CHECMATE program and engineering evaluations of susceptible systems.  ; Approximately 4000 feet of small bore pipe were planned to be replaced with upgraded material (SA 335-P11). Procedure PBM-UT-601 outlines the implementation of the FAC ultrasonic testing, including establishing NDE reference points for FAC examinations and analysis of data. The areas reviewed by the NRC were selected from the ESW system. The data reviewed was component selection, ultrasonic inspection, and data evaluation. In addition to the data review, the NRC ultrasonically examined three areas that were previously examined by the licensee, included in this examination were the following areas, pipe to elbow and elbow to tee upstream from Pump 2BP163. Results: The thickness reading taken by the NRC closely matched those taken by the licensee. In addition to those areas done by the licensee, the NRC did other spot examinations. No areas were identified being at or near the minimum thickness as required to be reported by the licensee's engineering department. 2.0 REVIEW 0F SITE NDE PROCEDURES MG MANUALS (73052) Peach Bottom Atomic Station submitted their updated second interval in-service inspection program to the NRC on July 7, 1990. .The items and areas planned to be examined in this program are in accordance with the Plant Technical Specifications and the 1980 Edition of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code up to and including Winter,1981 Addenda (80W81). I 1

   . -    _      _ _ - _ _ . . _ _ _ . _ _ - ~ _ _ _ . _ _ . _                _ _ _ _ . . _ _    - . _ _ _ _ . - _ _ _ _ . _ . -

7 The following licensee procedures were reviewed for compliance to the applicable Code, standards and specifications. General Electric (G. E.) Procedures Procedure Title Number /Rev. Qgig l Magnetic Particle Examination MT-PE-001, 2 06/10/93 l Procedure Liquid Penetrant Examination LP-PE-001, 4 06/15/93 Procedure l Manual Ultrasonic Examination UT-PE-002, 8 07/26/93

of Similar and Dissimilar Metal welds in piping systems Supplemental for Manual Ultrasonic UT-PE-002, 4 06/09/93 Examination of Dissimilar Metal Welds and Weld Butter Materials Procedure for Automated Ultrasonic UT-P8M-208V0, 0 08/25/94 Examination of Similar and Dissimilar Metal Piping Welds Procedure for Automated Ultrasonic UT-PBM-209VO, 0 08/25/94 Examination of Dissimilar Metal Nozzle to Safe End Welds Automated Ultrasonic Examination UT-PBM-503V1, 0 9/21/94 of the Shroud Assembly Welds PEC0 Eneray Procedures Certification of Personnel in MAG-CG-430, 1 06/01/94 Nondestructive Examination (NDE) Methods Ultrasonic Examination for F12-CG-406, 2 06/01/94 Erosion / Corrosion l Visual Examination of Pumps, MAG-CG-407, 2 06/01/94 Valves, Bolting and Component Supports Visual Examination of Reactor MAG-CG-408, 2 06/01/94 Vessel Internals Visual Examination of Power MAG-CG-401, 2 06/01/94 Plant Components, Welds, and j Equipment l

8 Liquid Penetrant Examination MAG-CG-402, 2 06/01/94 Magnetic Particle Examination MAG-CG-403, 2 06/01/94 Ultrasonic Examination of Raw MAG-CG-409, 0 12/13/93 water piping systems for Corrosioa Radiographic Interpretation MAG-CG-410, 2 06/01/94 Ultrasonic Examination of Welds MAG-CG'-411, 1 06/06/94 Manual Ultrasonic Measurements MAG-CG-412, 2 06/01/94 of Material Thickness Weld Crown Location and Marking MAG-CG-417, 0 12/13/93 Enytti: No violations or deviations were noted. The procedures appeared complete and functional. The following weakness was noted: Procedure MAG-CG-417, Rev. O, provides the steps for weld crown location and marking. The existing safety related plant welds examined during this inspection did not have weld centerline location or reference marks stamped, per this procedure. Article 111-4330 of the ASME Code Section XI requires a reference system to be marked on the piping to locate the weld centerline. Since this Code requirement was developed after the plant construction, the licensee has taken the position that placing reference marks on all piping welds that require surface or volumetric examination would constitute a backfit. The inspector noted that other licensees when faced with this Code revision have requested relief from the requirement to mark all applicable welds but have generally agreed to mark welds as they are ' examined as part of the ISI program. 3.0 MANAGEMENT MEETINGS Licensee management was informed of the scope and purpose of the inspection at the entrance meeting on September 26, 1994. The findings of the inspection were discussed with the licensee representatives during the course of the inspection and presented to licensee management at the exit meeting October 7, 1994. The licensee did not indicate that proprietary information was involved within the scope of this inspection, nor did the licensee object to any of the findings of the inspection. The following individuals were contacted: i

 ..   '.                                                                                                    j l          ~                                                                                                 l 9                                          ,

1 PECO Enerav l i

• T. Anderson Jr. NDE Specialist l
• J. Armstrong Sr. Mgr., Plant Engineering )

G. Budock Outage Planning  ; R. Ciemiewicz Engineering ISI Program )

• G. Edwards PBAS Plant Manager  ;

J. Giampietro NDE Technician i J. Hawkins NDE Specialist M. Harvatinovic Health Physics S. Kohibus Health Physics t

• D. Le Quia Maintenance Director
• 0. Limpias Mgr., Civil /Str Branch D. Morgan Q. V. Supervisor M. Lind Manager, MND Support

!

• G. Ruf . ISI, FAC Engineering

.

• R. Smith Experience Assessment l
• J. Stanley Component Engineering, ISI, FAC l
• M. Wagner Maintenance Supervisor l General Electric Comoany l

l E. Reczek Projects Manager l W. Miller Project NDE Level III U. S. Nuclear Reaulatory Commission W. Schmidt SRI, PBAS l

• R. Lorson RI, PBAS .

M. Modes Chief, Materials Section, RI l

• Denote those attending entrance and exit meeting.

The inspectors also contacted other administrative and technical personnel during the inspection. l l l  ! l i 1 l l l l - . l

!',*- . 10 l NRC NDE NOBILE LABORATORY TABLE No. 1 WELD ID. No. SYS NONDESTRUCTIVE TEST SHT.# 1 OR OR l ISO / DRAWING LIN CL RT UT PT NT l VT ACC REJ 6-B-10 FW 2 X X X X 6-A-2 FW 2 X X X X 6-A-9 FW 2 X X X X 6-A-13 FW 2 X X X X 1 12-10-09 RWCU 2 X X X X i 12-10-11 RWCU 2 X X X X l l 12-1-1 RWCU 2 X X X X 1 ( 12-1-1B RCWU 2 X X X X l l 12-1-1A RWCU 2 X X X X 12-1-1D RWCU 2 X X X X ! 12-14-5 RWCU 2 X X X X l 12-14-19 RWCU 2 X X X X 10-2HS6-1 RHR 2 X X X 10-2HS6-18 RHR 2 X X X 11-2CSIA12-22 RHR 2 X X X (*)2-BKA-8 RPV 1 X X 1 (*) EXAMINED WITH SAFT-UT SYSTEM l l l l I

lg,e s,s Riog d Iog 8[ I o UNITED STATES NUCLEAR REGULATORY COMMISSION Y WASHING ton, D. C. 20555

      %,...../

May 12,1980 l l MEMORANDUM FOR: B. H. Grier, Director, Region I J. P. O'Reilly, Director, Region II J. G. Keppler, Director, Region III i K. V. Seyfrit, Director, Region IV . R. H. Engelken, Director, Region V FROM: Norman C. Moseley, Director, Division of Reactor Operations Inspection, Office of Inspection and Enforcement

SUBJECT:

IE BULLETIN No. 80 CRACKING IN CORE SPRAY SPARGERS The subject IE Bulletin should be dispatched for action May 12, 1980 to all BWR's with aa operating license. The Bulletin should be sent for information to BWR facilities with a construction permit. The text purpose. of the Bulletin and draft letter to licensees are enclosed for this p ko a

                                                                        . Moseley, Director Divis n of Reactor Operations Inspection Office of Inspection and Enforcement

Enclosure:

1. Draft Transmittal Letter

2. IE Bulletin No. 80-13 l CONTACT: R. A. Hermann, IE i

49-28180 l l

(Draft letter to all GE power reactor facilities with a construction permit.) IE Bulletin No. 80-13 Addressee: The enclosed IE Bulletin No. 80- 131s forwarded to you for information. No written response is required. If you desire additional information regarding  ! this matter, please contact this office.  ! Sincerely, Signature (Regional Director)

Enclosures:

1. IE Bulletin No. 80-13 '

2. List of Recently Issued IE Bulletins j

4 e

6 IE Bu11 tin No. 80- 13 May 12, 1980 Page 2 of 3 The repair measures proposed were determined by the NRC to be adequate until the following refueling outage. The NRC evaluation stated that actions should be taken to refueling develop and install an improved replacement system at the following outage. Pilgrim Nuclear Power Station On January 31, 1980 the Boston Edison Company (BECo) informed the NRC that five indications in the upper core spray sparger and two indications on the lower core spray sparger at the Pilgrim Nuclear Power

• Station were identified during remote visual inservice inspections. The indications.were confirmed as cracks after hydrolasing and brush cleaning. The licensees evaluation indicated I that the sparger will retain structural integrity throughout the next cycle, although core spray flow distribution may be affected due to through-wall cracks. However, core spray flow delivery to the shroud interior would not be

    . expected to decrease. A loose parts analysis was presented which addressed (1) corrosion, (2) flow blockage, and (3) control rod interference.                 .,

To support power operation in Cycle 5 with the core spray sparger in its present condition, BEco has reanalyzed ECCS taking credit only for core spray reflood, taking no credit for core spray heat transfer. The submission by BEco is currently under review by the staff. The analysis is expected to cover a full spectrum of core spray failures. It is expected that the limiting condition will be the failure of recirculation suction line. A MAPLHGR limit reduction.will likely be imposed during Cycle 5 to compensate for the assump-tion of no core spray heat transfer. ,

                                                                                              )

Based on results from other sparger inspections and previous pipe cracking experience, cold work and sensitization during fabrication and installation stresses are considered to be the major factors in causing the observed cracks at the Pilgrim Station. The cracks are hypothesized to be initiated and propagated by intergranular stress corrosion (IGSCC). i A meeting was held with representatives from GE in Bethesda, Maryland on  ! March 13, 1980 to discuss core spray sparger cracking at BWRs. At the meeting ' GE provided the following information:

1. In February 1979, GE issued to BWR licensees Service Information Letter (SIL) No. 289 that recommended inspection of the core spray spargers for visual indications of cracking. To date, 19 of 21 plants inspected have no observed cracking. Cracks have been found at 2 facilities (Pilgrim and Oyster Creek).

2. The key contributors to IGSCC vary from plant-to plant, although stresses from cold work and sensitization during fabrication and installation are l considered prime factors leading to IGSCC at Pilgrim and Oyster Creek. '

Because the cause of cracking is not yet confirmed by metallurgical analysis, GE is developing tooling to extract sparger samples to verify 1 the postulated cracking mechanism.  !

l' IE Bulletin No. 80-13 Enclosure May 12,1980 RECENTLY ISSUED IE BULLETINS Bulletin Subject Date Issued Issued To No. 80-12 Decay Heat Removal System 5/9/80 Operability Each PWR with an OL 80-11 Masonry Wall Design 5/8/80 All power reactor facilities with an OL, except Trojan 80-10 Contamination of 5/6/80 All power reactor Nonradioactive System and facilities with an Resulting Potential for OL or CP Unmonitored, Uncontrolled , Release to Environment , 80-09 Hydramotor Actuator 4/17/80 All power reactor  ! Deficiencies operating facilities and I holders of power reactor construction permits 80-08 Examination of Containment 4/7/80 All power reactors with Liner Penetration Welds a CP and/or OL no later than April 7, 1980 80-07 BWR Jet Pump Assembly 4/4/80 All GE BWR-3 and Failure BWR-4 facilities with an OL 80-06 Engineered Safety Feature 3/13/80 All power reactor (ESF) Reset Controls facilities with an OL 80-05 Vacuum Condition Resulting 3/10/80 All PWR power reactor In Damage To Chemical Volume facilities holding Control System (CVCS) Holdup OLs and to those with Tanks a CP 79-018 Environmental Qualification 2/29/80 All power reactor of Class IE Equipment facilities with an OL 80-04 Analysis of a PWR Main 2/8/80 All PWR reactor facilities Steam Line Break With holding OLs and to those Continued Feedwater nearing licensing Addition 80-03 Loss of Charcoal From 2/6/80 All holders of Power Standard Type II, 2 Inch, Reactor OLs and cps Tray Adsorber Cells

  .    .        .    .       . . . _ _         _ ~._             .     .   . . _ _        -  _      _

T+c 193 7/ 4* 2+*E /Af

      /         o g       7                          UNITED STATES g                   NUCLEAR REGULATORY COMMISSION

[ WASHINGTON, D. C. 20555 g 8 June 10, 1982 k..... . Docket No. 50-277 Mr. Edward G. Bauer, Jr. Vice President and General Counsel Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101

Dear Mr. Bauer:

By letter dated May 11, 1982, and as supplemented on June 4,1982, you informed us of a crack in the "B" core spray sparger of Peach Bottom Unit 2. The crack was discovered through an inservice inspection required by IE Bulletin 80-13. " Cracking in Core Spray Spargers." You stated that an svaluation of the crack concluded that no modifications were required to ensure continued safe operation of the reactor. How-ever, you decided to install a clamp at the crack location to provide further assurance of core spray sparger operability and safe reactor operation. M addition, you reanalyzed the effect of complete severence of the sparger in regard to the design basis accident. Your June 4 1982 letter enclosed a General Electric Report entitled, " Core Spray Sparger Crack Analysis at Peach Bottom Atcaic Power Station Unit 2," NED0-22139, May 1982. The GE Report concluded that no loadings have been identified which could result in stressei that would cause the spargers to break during normal plant operation, transients, or postulated loss-of-coolant accidents, without the installation of a clamp. We concur in this conclusion. Section 4 of the GE Report and the May 11, 1982 letter provided the results of a reanalyses of the loss-of-coolant accident. The analysis shows for the limiting case of a single failure of one core spray with the remaining sparger assumed to be one with the cracks, that the peak clad temperatures in the fuel would remain below the 10CFR 50 Appendix K limits without changing the limits of maximum average planar linear heat generation rates within the bounds of the curves of Peach Bottom 2-Cyc1t. 6, the ' current core loading. We concur in these conglusions'.- Based on the above we conclude that operation of Peach Bottom Unit 2 with a cracked, but clamped "B" core spray sparger will not result in e an unsafe operating condition. , Sincerely, n A (3 J' ]< Jol)EF. Stolz, Chief a 4)perating Reactors Branch #4 Division of Licensing cc: See next page f M O$N-N

 ..,d,-
                                                         .~       . - - .        . . . . . , . . . . . . . . . .
                                                                                                                        ......_..a.u.....

l TE 493 7 / 4 REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS) [ ACCESSION NBR 8205070269 DOC,0 ATE: 82/04/29 NOTARIZED: NO DOCKET # ( FACIL:50-277 Peach Bottom Atomic Power Station, Unit 2, Philadelph 05000277 AUTH.NAME AUTHOR AFFILIATION GALLAGhER,J.W. Philadelphia Electric Co. RECIP.NAME RECIPIENT AFFILIATION HAYNES,R.C. Reg' ion 1, Office of Direct'r o

SUBJECT:

Responds to IE dulletin 80-13, " Cracking,in Core Spray . Spargers." Present plans direct towards installation of clamp which will assure positioning of sparger arm. DISTRIBUTION CODE: IE11S COPIES RECEIVED LTR JL ENCL'JC) SIZES.shI..._ [ TITLE: Bulletin Response (50 DKT) , l i NOTES: RECIPIENT COPIES RECIPIENT COPIES l 10 CODE /NANE LTTR ENCL ID_CQQ NAFE LTTR ENCL ' i ORB #4 BC 1 ([' _FAIRTTLE, . 1 INTERNAL AE00 1 IE FILE 01 1 NRH/DL/0RAB l' EXTERNAL: LPOR 1 NRC POR 1 NSIC 1 j . NTIS 1 1 1 , J i O TOTAL NUMBER OF COPIES REGUIRED: LTTR 9 ENCL sf

                                                                    - - - . . . .      a. .   =_a      = :: .. .. .             .. , 2 ...:.u ..a

} . . r . ( . ,

f . . Li ' 500'F) is necessary for the recovery phenomenon, and such thermal treatment was not practical for the spargers,'nor deemed necessary. The most significant influence of cold work is in the transformation of austenite to martensite phases through deformation. Martensite, if present in suf ficient quantity, can assist in recrystallization of the material upon subsequent thermal treatment. The strain energy induced in'the lattice promotes recrystallization. The result of recrystallization is migration of grain boundaries away f rom chromium depleted regions, with attendant benefits in reducing sensitization. However, the presence of martensite increases the tendency for carbide precipitation and local chromium depletion during subsequent veld sensitization. A wider HAZ can result from welding stainless steel with prior cold work-induced martensite. If suf ficient cold work is 2-14 l

l

             .                                                                                            l NEDO-22139 I

present, transgranular cracking can occur in .orygenated water environments

         ' with or without subsequent sensitization.

Environmental tests conducted on tensile, bent beam and pressurized tube specimens are illustrated in Figures 2-14 through 2-17 (which are based on information from References 2-3 and 2-4).

                                                                                      "***# I #

In Figure 2-14 it can be seen that the time to failure in 0.2 ppm 02 l sensitized and cold-worked and sensitized material varies with stress. Speci-mens tested at cold-worked-plus-sensitized conditions (at higher stresses)  ! produced failure times (by IGSCC) comparable to samples which contained no l work prior to sensitization. Cold-worked samples without. subsequent sensitiza-tion, tested at comparable stresses, did not fail. IGSCC failures could be

        - induced at very high stressesin cold-work /nb-sensitiration samples, as-illustrated in Figure 2-15.
                                       ~               '                                             -
            'If the data from Figures,2-14 and' 2'-15' are plotted on a basis' normalized by the---

material yield strength, a more clear picture is formed of the results of deflection-induced stresses in stainless steel (Figure 2-16). Material cold i worked to various levels and subsequently sensitized can undergo stress corro-sion at substantially lower percentages of the material yield strength, with cracking as low as 80% yo , in quarter-hard stainless steel (furnace sensitized). I l An equivalence must be established between plastic strain during sparger arm The yield strengths forming and'the cold-work condition of the test specimens. of specimens receiving 5, 8, and 15% cold work are illustrated in Figure 2-17. The stress-strain curves for the same heat of material without prior cold work indicate the amount of plastic strain necessary to create a comparable yield f stress to the uniformly cold-worked specimens. Thus, 2.1% plastic strain cal-culated for arm bending correspond.s to approximately 1% cold work and stresses 1 near yield may or may not result in cracking (data are insufficiently clear). l The strain concentration from localized bending, if a factor of 4 is considered, would be comparable to 5% cold work. A reduction of the cracking threshold to 0.8 e and cracking under residual and installation stresses could occur. 2-15 i

NEDO-22139 . 2.6.3 Conclusions of Sparger Cracking Core spray sparger cracking at the Peach Bottom-2 plant can be il t hypothesized b the influence of weld sensitization or prior sensitization of the arm maSources er a and subsequent cold work of the arms during forming and installation. ld of stress for IGSCC are dependent on residual stresses from arm bending, we . residual stresses, and deflection during installation. The principal factors suspected of causing cracking are considered to be hig The absence of one or several key factors variable from one plant to another. may explain the lack of reported indications in the majority of the BWR operating plants inspected to date (May 1982). 2.7 CRACK ARREST ASSESSMENT the following sources of stress In assessing the possibility of crack arrest, are considered:

1. Stress due to pressure, mechanical loads and thermal gradients.

These stresses have been shown to be negligible and are not considered in the crack growth assessment. these are displacement controlled

2. Stresses due to bracket restraints (secondary) stresses and would be expected to relax as the crack propagates.

as the crack propagates into a

3. Residual stress due to fabrication:

region of compression, the stress intensity factor can be expected to decrease, thereby resulting in arrest. weld residual stresses at the T-box - sparger

4. Weld residual stress: These stresses are likely to welds would influence crack propagation.

vary circumferential1y and also relax as the cracks become larger.

5. Stresses due to vibration are assumed to be negligible.

2-16

• NEDO-22139

( . In considering crack arrest, the stresses due to bracket restraint and ' i the f abrication residual stress are significant and are evaluated in L de t ail . -- -_. - , , , , _ _ , _ _ _ _ _ _ , 1 1 l 2.7.1 Stresses Due to Bracket Restraint Stresses due to bracket restraint are governed by the applied displacement i and the compliance of the pipe. Since the displacement is fixed, the This is compliance change with crack growth could lead to crack arrest. comparable to crack arrest in a bolt-loaded wedge-opening-loading (WOL) l j' specimen in stress corrosion tests. Figure 2-18 shows the variation of . The compliance co'apliance with crack length for a pipe subjected to bending.

  *" *

• was determined-using-the relationship betveg.rg the ..s. train..ene.r.gy r..elease - - - .. .. .. - . . _ ..

rate G and the compliance change per unit area of crack extension dc/dA l (Reference 2-5). For the cracks in the sparger, L/d is expected to be l in the range of-0-< L/D-<- 40...F4gure.2-13. showothat,, when.more ,than,30% , , _ [' ~ " of the pipe is cracked, the compliance of the pipe increases by a factor p ! of 10. Therefore, for the given initial displacement, the stress in the

                                         ~                           ~                                -

sparger and tee appl 1ed' stress iritettaity -f actor-would-decrease by.a. factor.._ f of 10 when more than 30% of the pipe circumference is cracked. Clearly, the the crack length exceeds this value, the restraint stresses become i negligible and crack arrest is expected. 4 i 2.7.2 Fabrication Residual Stress i-j The residual stresses due to fabrication vary around the circumference, t and a precise calculation of the stress intensity is not possible. Nevertheless, a conservative representation of the stress is used to l calculate the stress intensity f actor. The assumptions made are as 4 follows: i j l 1. The crack in the sparger is modeled as a through crack in an infinite plate. i i 2-17

      + - , ,             m-        .                      -,              . . - , , - - ,                       m -        --       - , . -       - - . _ - -             -       r     -w

NEDO-22139

                 '2.

It is assumed that the tensile stress (a) is uniform and is applied - on the crack face over a length (2b). (Later this will be conserva-tively taken as 25% of the circumference.) The remaining portion of the crack is assumed to be subjected to a 3. compressive stress, which is half the tensile stress (Figure 2-19).

4. The ' crack length '(2a) for which the combined stress intensity f actor reduces to zero is calculated.

The stress intensity factor due to the tensile stress can be shown to be: l K

                        ""8        "=   #"
                                           - sin ~b\a/

I g l The stress intensity f actor due to the compressive stress e/2 is given by: I

                                                                    ~

1 - sin compression 1 ,6-2(o/2) a <[2 l Setting Ktens 7

                                      ,geomp = 0, we get
                                                       ~1 sin                =f       - sin i
                          ~
                                        =f or, sin                                                                                                      I or, b = 0.5a and the remaining If we assume b = 25% of the circumference is under tension o y                                              1 portion of the crack is under compressive stress (equal to half the tensile)                                !

stress), the applied stress intensity factor bec'omes zero when the crack length Thus, even under extremely conservative j is equal to 50% of the circumference. assumptions, crack arrest is expected. l l I 2-18 l l I

1 .i ., 3 4

           -                                                                            NEDO-22139 i

l 2.7.3 Conclusions on Crack Arrest l Based on the above material, the following conclusions may be drawn: 1 1 a 1

1. Since the applied loading is predominantly displacement controlled,

]I the stresses can be expected to relax as the cracks grow. Crack 5 arrest is therefore expected. 1 1 1 I 4

2. The residual stresses due to fabrication vary from tension to compres-I cion. As the cracks propagate into regions of compressive stress, the K value reduces to zero. Even for extremely conservative assump-

! tions, crack arrest can be shown for a 50% circumferential crack. J l 3. The above conclusions are valid as long as there is no stress cycling

                                                                                                   -   -~

i duetovibratilA~ie.g., flow-inducedv'ib' ration). l l l 2.8 STRUCTURAL INTEGRITY WITH CRACKS i ' From the discussion of the potential stresses in the core spray sparger (Sections 2.5 and 2.6), it is concluded that only deflection limited secondary stresses approach 25% of the material yield strength (except for self equili- i brating thermal stresses). If a 360* throughwall crack is postulated at any location on any sparger arm, the remaining stresses will not produce a failure at any other location on the sparger. The AP stress and the stress resulting-from an axial load in the pipe due to bracket friction are proport'ional to the cross-sectional area of the pipe. The load from AP and friction was found to be <1000 lbs. Assuming a yield strength of 30,000 psi at core spray flow temperature, an area of less than 0.033 in.2 is required to maintain continuity. This area is much less than the original pipe metal area of 3.17 in.2 The . bending type stresses are all deflection limited secondary stresses. The discussion in Section 2.7 shows that cracks are expected to arrest, since the driving stress will be relieved. The bending loads may, however, in a worst case, cause an existing crack to open up by an additional 0.005 in., assuming the existing crack has progressed 360'. This is a geometry limited condition.

                                                                                             '2-19

NEDO-22139 p It is concluded that no loadings have been identified which could result in stresses that would cause the spargers to break during normal plant operation,

• transients, or postulated loss-of-coolant accidents.

2.9 REFERENCES

2-1. Hopkins, Design Analysis of Shafts and Beams, McGraw-Hill Book Company. 2-2. H. H. Klepfer, et al, " Investigation of Cause of Cracking in Austenitic l Stainless Steel Piping," NED0-21000-1, July 1978. 2-3. A. E. Pickett and R. G. Sim, "The Effect of Stress and Cold Work on Intergranular Stress Corrosion", Materials Protection and Performance, l Vol. 12, No. 6 June 1973. I

                                                                                                                                    - -- t 2-4. G. M. Gordon and R. E. Blood. " Reactor Structural Materials Environmental Exposure Program", Symposium on Materials Performance in Operating Nuclear Systems, Ames Laboratory, Ames, Iowa,' August 28-30, 1973.

2-5. E. Kiss, J. D. Heald, D. A. Hale, " Low Cycle Fatigue of Prototype Piping", GEAP-10135, January 1970. I 1 1 l 1 2-20 ) l

l NEDO-22139 Table 2-1 POSSIBLE CAUSES OF CRACKIh. Possible Cause Evidence Location Sensitization by Welding. -ation of cracks

1. Sparger Arm i Near T-Box ;1 mate 5% Cold Work I Cold Work Followed by  !

Weld Sensitization ..sar T-Box I Weld Residual Stresses iscation of cracks Fatigue (thermally Induced) '.T's are Low I Fatigue (Flow-Induced Vibration) Amplitudes are Limited 1 l I Sone *

2. Sparger Arms Sensitization from Fabrication 1 Cold Work Followed by Pipe Bend Forming *,

Away from No Evidence of , T-Box Sensitization Sensitization l Local Heavy Cold Work None*

                   ,                                                                         ~~
                                                                  'Same as in 1. Above ~

Fatigue

• Sensitization and cold work state of spargers not yet known.

2-21

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N)E l . i 1 , lW Figure 2-2. Core Spray Sparger - Plan View j 2-23 1

NEDO-22139 , lP N/ ,~ l UNE5_ j N i

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                                         )

I l I - 192 in.mcIP'6Picii l 'f it'O LOWER SPAplGER I I top GUIDE i Figure 2-3. Sparger to Shroud Attachment Method 2-24

        -. _ . . ._           = .. .              .   -- .- . - .-.
    -                      NEDO-22139 I

1 0 4 , g 7

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f,h p,

                       /
                       /

6 Figure 2-4. Sparger Nozzles 2-25

NEDO-22139 sHmouc

                 // D[

LotoTosM ovo I )i h [ oNE E El i n eyh Lowsm smAcxt_T Figure 2-5. Sparger Support Method 2-26

   -                                             -                       ._ - .           -. - l

 . -_.       . . .__    . _ . . _ . . . . _ _ _ . _.  =_m.        . - .      . ._._,__._._____..m        .        . - _ . . _ _ . . . _ _ _

d NEDO-22139 4 k i

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l } e i 105.75 gn,

                                                                        '3'
                                                                        ~
         .                                                              *T.
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Figure 2-6. Pipe Bending Method 1 2-27

NEDO-22139 .

                               ' INITIAL CONFIGUR ATION T                                                                                        'M, d'                             .

LOAD APPLICATION DU. RING. P ABRICATION l l 1 k ,

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                      , FINAL DEPORMED SHAPE AFTER LOAD REMOVAL Figure 2-7. Sequence of Events Leading to the Residual Stress Distribution 1

l 2-28 i

   . _ . _     -. ~ - - . , . - _ .                . . .           ~ _ . . - - - - - _ . _ - . . - . . _ . - . _ . - . - _ . - . . . . - . . _ . - . .                                ..~ ~ .. .. . . - - . _ _ . -

q t NEDO-22139 l f 1 e l l l ROOM TEMP (74*Pl

                       ~                                            _.                                                              - _ .                       .,_

3 e j i It , j I I' f

                                      .y  -

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                                                          'y

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i i a *, -. i 1 i

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) ey = 0.0007 j 3

 ;                                                                                                                                                                     E= 25.8n10 hi 3

i E=T 0.26 m 10 ksi J 4 I a i e# E

 $                                                                                                                                             M 1

i Figure 2-8. Bilinear Stress-Strain Curves for Type-304 j Stainless Steel i f

 <                                                                                                                         2-29

6 NED0-22139 . t 1 p \;. *.*- =j I e,) [e e,

• E, (so h =

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                                     \

STRAIN  ! i loisTmisuTioN 1 I Figure 2-9. Stress and Strain Distribution in the Pipe Under Applied Moment 2-30

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FUL RE'S4$UAL~~ ~ sinEss pesTatsuitoes Figure 2-11. Resultant Residual Stress Distribution After Fabrication

• 3

t 9

       ~                                          NEDO-22139 i

e INSTALLATIONJ AGI AL MISM ATCH DIP P ER ENft'Attwt LO

                                          '$NRINKAGE AT T.toX 5
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                                                   FOR SHRINK = 1/8 In.                                         1 e AT T BOX a 21 kai (ELASTIC)

ASSUME R1 = 104,75 R2 = 105.15 _, UNIFORM FORMING l e a 1.0%  ; e a 38 ksi (FROM o - e CURVE) 1 1 Q iPi 4 42_ f M 1 Tf WI i 1

                                                                                       =--e.

l Figure 2-12. Postulated Installation Stresses 2-33

     .. - - -       _ -                .       _.      . ____ -                   - . _. - . -. . - -= - ._ - . -                                            . _ . . . . - . _ - - -

i NEDO-22139 i . 1 ! , OsCREASES

                 ""TeiB TAggf""   *
                                                                                                                                         .suscarTisitiTY suscseTisiuTY                                                                                                             TO C8 AC"'NO

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                                                                                                                       ~lMPOSED LOADS
                                                                                                                                           - j L       _ _ . _ _ . _ _                _ _      . oma _                         _ __ -

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                                                                                                                                                         ', RECOVERY
                                                                                                                                ,     INC.R EASES
                                                                                'gLD WORK
                                                                                         .M I
                                                                                                                                    ' CH_ ROM LUM ACTivlTY*~
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s - I PRC"DVcTS- ' RTENSlTE """""~~~~~*""~~l 1 1 1 Y . I AIDS RECRYSTALLIZATION iP 6GSCC iP SENSITIZED._ 8 CW > 1S%~ AND ME AT TREATED I

                            . APTER MODEP ATE COLD ,                                                         -

V MIM'TEMPIRATURE~~~*~ ' WOR K ISTR ESS R EQUIR EDI

                  +              u c. =: :- ;
                                                                .=: - _               , ,,,,,,                                                              -

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                       . COLD WORK tsTR,,         E.SS REQUI_ RED)   .

i , I 4 Figure 2-13. Effects of Cold Work on IGSCC of Type-304 j j Stainless Steel 2-34 1 4

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NEDO-22139

3. LOST PARTS ANALYSIS o

3.1 INTRODUCTION

Based on the structural analysis given in Section 2, it is expected that the Peach Bottom core spray sparger willnot break and result in loose pieces in the reactor. However, an evaluation of the possible consequences of a poten-tial loose piece is presented in this section. 3.2 LOOSE PIECE DESCRIPTION Since a piece has not been lost, it cannot be uniquely described. Three different types of loose pieces are postulated in Section 3.4.2: (1) a sec-tion of sparger pipe; (2) an outlet nozzle; and (3) a small piece of the sparger. The entire sparger is fabricated vf Type-304-e+einlese et=at 3.3 SAFETY CONCERN The following safety concerns are addressed in this safety analysis: 1

1. Potential for corrosion or other chemicai reaction to reactor materials.

2. Potential for fuel bundle flow blockage and subsequent fuel damage.

3. Potential for interference with control rod operation.

3.4 SAFETY EVALUATION l The above safety concerns for the postulated loose pieces are addressed in this section. The effect of these concerns on safe reactor operation is also addressed. 3.4.1 General Description The core spray spargers are attached to the inside of the core shroud (Figure 3-1) in the upper plenum. For a piece of the sparger to reach and 3-1

NEDO-22139 potentially block the inlet of a fuel assembly, it would have to be carried out of the upper plenum and pass down into the lower plenum. To accomplish this, it would have to be carried by the fluid flow in the upper p'lenum up through the steam separators then outward to the downcomer annulus, then through the jet pump nozzle into the lower plenum, then make a 180' turn and be carried upward to the fuel assembly inlet orifices. A part of the sparger cannot reach the fuel assembly inlet, orifices by falling down inside the core shroud For a part as the core support plate and the loaded core will prevent this. of the core spray sparger to reach a control rod, it must first traverse the upper plenum from the outer region of the shroud toward the center, which is unlikely, then fall through the restrictive passage between two fuel channels. Since all parts of the core spray sparger are designed for in-reactor service, there is no possibility that any loose part will cause any corrosion or other . chemical reaction to any reactor material. 3.4.2 Postulated Loose Pieces 3.4.2.1 Sparger Pipe The sparger pipe is 4-in. Schedule ,40 pipe and is attached to the core shroud at six locations (T-box plus five brackets) , The maximum span between supports In order to is about 38-1/2', which corresponds to approximately 71 inches. generate a loose piece of pipe, two throughwall cracks would have to propagate 360' around the sparger. The weight of the largest pipe segment would be approximately 90 lb. (1) the top A pipe segment could come to rest in any of three locations: surf ace of the top guide outboard of the fuel assemblies; (2) the top surface the to; surface of the fuel assembly handles; or (3) in an unlikely event, of the core plate. In all three of these locations, the flow velocity is Therefore, it will remain low and insufficient to lift a segment of the pipe. at one of these locations (see Appendix C for flow velocity calculations). f 3-2

NEDO-22139

 ,    A 90-lb piece of pipe which f alls from the core spray sparger will not harm the core plate, top guide or fuel assembly handles, since these components are designed for much larger loads.                                     .

Since the pipe cannot be lifted by the flow and since the pipe cannot fit through either the steam separator or the jet pump, it will not cause any flow blockage at the fuel inlet orifice. Since the pipe is too large to fit between fuel channels, it will not cause any interference with control rod operations. 3.4.2.2 Spray Nozzle Each spray nozzle consists of two 1-in. elbows fabricated of Type-304 stain-less steel, which are welded to the sparger. In order to generate a loose nozzle,a throughwall crack would have to propagate 360* around the nozzle. The weight of each nozzle assembly is approximately 1-3/4 lb. A loose nozzle would most likely come to rest on the top surface of the core plate or on the top surface of the top guide. The flow velocities in these regions are insuf ficient to lif t the nozzle, thus, it will remain at one of the above mentioned locations. i l fit Since the nozzle cannot be lif ted by the flow and since the nozzle cannot through the steam separator, it will not cause any flow blockage at the fuel assembly inlet orifices. The nozzle is too large to fit between two fuel channels; thus, it cannot cause any control rod interferences. 3.4.2.3 Small Pieces A small piece of the sparger could become loose if both longitudinal and cir-cumferential throughwall cracking occurred. A small piece could be lifted by the flow if it maintained an orientation with its maximum projected area perpendicular to the flow. Due to flow turbulence and nonsymmetry of the loose part, the part will tend to rotate so that the minimum projected area will be perpendicular to the flow. With this orientation, all parts with a length of greater than approximately 0.4 in. will sink (Figure 3-7 of Refer-ence 3-1). Thus, most pieces will not be carried by the flow toward the 3-3

NED0-22139

                                ~

steam separator. However, in the unlikely event that a piece reaches the

• steam separator, it would have to pass through the steam separator turning l vane (Figure 3-2). The turning vane has eight curved vanes. The outlet l

( of each vane overlaps the inlet of the adjacent vane. The longest straight 1 piece that can fit through the turning vane is approximately 6 inches long and it must be oriented with the long dimension in the vertical direction. The largest piece that can fit through the turning vane with its long dimen-sion in a horizontal plane is shown in Figure 3-3. It is very unlikely that the flow velocities would carry either of these maximum sized pieces through the turning vane. After passing through the turning vace, the fluid momentum is reduced as tha water is removed. At the separator exit, the fluid is almost entirely steam. A typical water content is 1 weight percent. Thus, it is very unlikely that any piece could be car-ried out of the separator by the steam. If any piece were carried through the 1 i separator by the steam, then it could be carried into the downcomer annulus, 1 through the je*. pump and enter the lower plenum. A piece that entered the i lower plenum would most likely be driven by jet pump flow to the bottom of the reactor pressure vessel where it would most likely remain. However, per Reference 3-1, a small piece could be carried by the flow up to the flow inlet orifices. The orifice sizes are 1.244, 1.469 and 2.211 inches. It is extremely unlikely for a piece larger than the 1.244-in. orifice and essentially impossible for a piece larger than the 2.211-in. orifice to be carried through the steam separator. The outside diameter of the sparger is 4.5 in., while the fuel inlet orifices are slightly recessed relative to the surface of the control rod guide tubes (Figure 3-4), which have an outside I diameter of 10-7/8 inches. Due to the different radii of curvature, flow would be able to enter the fuel assemblies. Thus, unacceptable flow blockage as defined by Reference 3-1 would require that more than one loose piece be carried to the same inlet orifice. This is based on the size of the piece (s) that, in a highly unlikely circumstance, have the potential of reaching the vessel lower plenum. The probability of unacceptable flow blockage of any f i fuel orifice is judged to be insignificant. I 3-4 i - ___

NEDO-22139 t The flow velocities near the sparger are lower than those above the fuel assemblies. Thus, it is unlikely that a small piece would be carried over the fuel assemblies. If the piece were carried over the fuel asse,mblies and L then rotated-so that- the flow-could-no longer carry it, the-piece could _f all__. . . ___ _ . . on top of a fuel assembly or between fuel assemblies. Figure 3-5 shows a typical unit cell of four fuel assemblies and one control rod. The control rod moves in the gap between the fuel channels. The gap between fuel channels is 0.75 inch. The length of the gap between the channel spacer and the channel fastener is 2.3 inches. Thus, any piece larger than l 2.3 in by 0.75 in. cannot cause control rod interference. The control rod thick. ness is 0.312 in. and the diameter of the control rod rollers is 0.520 inches. Thus, pieces smaller than 0.334 in, will fall past the control rod without causing any interference. A piece of precisely the right size could be in contact with the control rod and one or two fuel channels. Such a piece might be detected during the normal control rod exercising. The rods are inserted one notch and withdrawn one notch each day. It is also possible, though unlikely, that a piece might wedge between two fuel channels above the f control rod and thus not be detected by normal control rod operation. If the rod were to be inserted, the control rod mechanism has enough force to lift l one fuel assembly with the reactor at normal operating pressure. If the fuel assembly were lif ted 1 or 2 inches, it would be able to move horizontally at both the bottom and the top, thus most likely relieving any interference. The rod would then insert and the fuel assembly would fall back into place. Thus, it is very unlikely that any control rod will fail to insert. One of the licensing bases of the reactor is that the highest worth control rods can be fully stuck out and the reactor can be safely shut down. Thus, unacceptable control rod interference will require multiple precisely-sized pieces interfering simultaneously with control rods that are in close prox-imity to each other. The probability of this is judged to be insignificant.

3.5 CONCLUSION

i i The probability for unacceptable corrosion or other chemical reaction due to l l a loose piece is zero. The potential for unacceptable flow blockage of a 3-5

NEDO-22139 fuel assembly is essentially zero. The potential for unacceptable control rod interference is essentially zero. 3.6 REFERENCE 3-1 " Consequences of a Postulated Flow Blockage Incident in a Boiling Water Reactor", October 1977 (NEDO-10174, Rev.1). 3-6

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s .,;.0. PsECE . :. L3.g.v e , . l.

                                                                                                               ;:n r..y t. v ; .;. vic g.g
                                             ,s.ruEL PEtt.ETs                                                                                                                        .,

14 END PLUS

                                                                                                                                        * *j !                                         to               ,,G 1 '                                        %

i

                                                                                                                                                                                                                                                                  ,        'I %, . 6.    .,.y .      -

l ,, j sp{.y,ki. '.., .d\, .. d 5.- O. . '.' ,4

• f.,,. ,;

ss. CHANNEL 4 sPACEm h n g ;4IYf,:1.. - a  ; ,

                                                                                                                                                                                                                                                                       .           r     3f ::       i                    .

{ N.9 , :s.. .. f.. f ,)

                                                                                                                                                                                        ' 8g u,

16. PLENUM o'

{t .i :. .4

                                                                                                                                     .;

• w - d< -

                                                                                                                                                                                                                                                                                                                       ,A j                                                                           sPaiNo                                                                                                             o                 -

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                                                                                                                                    . h' . . .                                                                                                                                      *                              //
                                                                                                                 ?l% J,s                                                 >                                                                          .                                                          ,,

l s.%k : q *% -c a , i GENER AL $ ELECTRIC l. h.

n. .

h '1}: -

                                                                                                                                                                                                                                                 -w4 hf ,                                       ,k,n'l:'.'-l h                                     ri .l.

j Figure 3-5. Fuel Assemblies and Control Rod Module l 3-11/3-12

{ ' l

           -                                               NEDO-22139 I

! I

4. LOSS-OF-COOLANT ACCIDENT ANALYSIS WITH NONUNIFORM SPRAY IN ONE SPRAY SPARGER l 4.1 DiTRODUCTION- ~

                                                                 -       -         -   .-                              . . _ _ _ _ (

This section describes the methods used to evaluate the MAPLHGR requirements to meet 10CFR50 Appendix K for the Peach Bottom Reload 5, Cycle 6, assuming no The inputs credit for core spray cooling from the cracked core spray sparger. to the approved 10CFR50 Appendix K computer codes are discussed in Section 4.2; the general sensitivity of the loss-of-coolant accident analysis (LOCA) results 1 I 1 to the spray cooling is discussed in Section 4.3; the results of the analysis are given in Section 4.4 and the conclusions are presented in Section 4.5. I I 4.2 INPUT TO THE LOCA ANALYSIS !~~~ ~Th'e~approvsd Viiirdiotis 'o f ' SAFE--1tEFLOOD, and-CHAST-E -codes-were used -to evaluate.. ...._.. ._ the impact of a cracked core spray sparger in Peach Bottom-2.

 - - - - -       The potenttal..ef fects -of cracks. in one _cor.e. spray, sparger. is_t,o cause nonuniform               , _ _ ,

If the second sparger is injecting flow spray distribution from the sparger. (i.e., the other core spray system is operable), the postulated effect could only reduce the amount of spray flow to the hot fuel assembly by the contribu-tion from one sparger. This ef fect is conservatively modeled by setting the spray heat transfer coef ficients in the CHASTE heatup code to one-half of their Appendix K values. This is the same assumption used in standard Appendix K analysis to model a core spray system out of service (Reference 4-1). If one core spray system is rendered inoperable due to the assumed single f ailure per Appendix K, the remaining sparger may be assumed to be the one with cracks. The bounding effect (the assumed lors of all spray to the hot fuel assembly) can then be represented by setting the spray heat transfer coefficients in CHASTE to zero. Therefore, in summary, the effect of cracks in one sparger is represented conservatively in this calculation by setting the spray heat transfer coef ficients to zero or to one half their standard value, depending on the single failure analyzed. l 4-1 l

l e- l

                                               .;9                                                       ,

l

                                                                     #ellowing This representation is very conserv.n'. = as discussed 1-paragraphs.

Counter current flow limiting (CCFL) is the phenomenon c .:ted with a steam

ich, in this updraf t limiting the downflow of water through a flow pr case, is the fuel assembly. The steam updraft in the fe - embly (due to rods subsequently) flashing during blowdown and to spray evaporation on the

er to an amount can, under certain conditions, limit the downflow of sp 1 smaller than the spray injection rate in the upper plen. .is CCFL is a

                                                                  ;ccause subcooling function of the subcooling of 'the water in the upper pit can quench the steam updraft and cause the CCFL to "breu...c.en," eliminating the " holdup" of the coolant downflow.

l Currently-approved Appendix K LOCA models assume saturated water CCFL condi-tions and conservatively ignore the inventory buildup of coolant in the upper l plenum. Recent large-scale tests confirm that the CCFL " breakdown" can occur soon af ter spray initiation, causing downflow of the upper plenum inventory and rapid reflooding of the core. Following this, a residual pool of ' Vater remains in the upper plenum, ensuring uniform delivery of coolant to the individual fuel bundles. The present core reflood time from Appendix K models does not model CCFL breakdown or the residual pool in the upper plenum. The effects of saturated CCFL modeled in the REFLOOD model produce an overly conservative estimate of the core reflooding time. If a crack, or cracks, forms in one sparger to the extent that the flow rate through the spray nozzles is reduced, then more injection will occur at the core periphery which will most likely cause localized subcooling and CCFL breakdown. This would reduce the reflooding time for Peach Bottom-2 up to 100 seconds from the value calculated with the standard Appendix K models resulting in PCTs up to 700*F lower. On the other hand, if no CCFL breakdown occurs, the upper plenum inventory builds up rapidly and ensures no reduction in coolant delivery from the core spray sparger system to the bundle and subsequently no degradation in cooling heat transfer. In addition to the above conservatisms, the 1973 ANS + 20% decay heat correla-tion was used in the analysis per Appendix K. The technical community at this 4-2

NEDO-22139 r time recognizes that the subsequent 1979 ANS decay heat correlation provides a more realistic basis for evaluating ECCS performance. This decay heat correla-tion would further reduce calculated steaming rates and CCFL effects, as well as ' the core heatup rate, which would reduce the calculated PCT an additional 200* to 400*F. 4 4.3 SENSITIVITY OF LOCA ANALYSIS TO NON UNIFORM SPRAY 4 For the Peach Bottom plant, there are no single f ailures for any break location (other than a core spray line break) that can render both core spray systems 1 I l inoperable. For core spray line break, there are always at least two low pres-For medium and large sure ECCS pumps available, ensuring timely reflooding. j break sizes (rhich depressurize relatively f ast), the most limiting f ailures are those that result in the least number of emergency core cooling system l ' (ECCS) pumps remaining operable.  ! i 1 The two single-failure candidates that are potentially limiting for medium to large' break sizes are: _ A. Diesel Cenerator Failure - 1 core spray (LPCS) + 1 Low Pressure Coolant Injection (LPCI) + HPCI + the ADS operable 4 B. LPCI Injection Valve Failure - 2 core spray + HPCI + the ADS operable Since the HPCI (High Pressure Coolant Injection) is steam turbine powered, it is not a significant contributor to mitigating medium to l'arge breaks. Also, since the function of the ADS (Automatic Depressurization System) is to depressurize the reactor as a backup to the HPCI, it contributes little toward mitigating medium and large break LOCAs. Therefore, failure candidates A and B each results in a dependence on only two ECCS. Per the Reload 5 analysis, failure candidate B (LPCI Injection valve failure) the fuel assembly is limiting because of the conservative modeling of CCFL at upper tie plates, which limits the downflow from the core spray systems and prolongs reactor reflooding. 4-3

    -. -. ~_                   - . - --           .- - ~ -.-           -  _. -.-   -.-       - . - . . . . . - . -

a NEDO-22139 i These two single-failure candidates were re-examined for larger breaks assum-

• ing a cracked spray sparger as described in Section 4.2. The limiting single failure, break size and location does not change, since the calculated core uncovery and recovery times and the reactor depressurization rates do not change with the methods described in Section 4.2.

For smaller break sizes, the limiting single failure is the high-pressure ECCS (HPCI), since the LOCA transient is a high pressure transient that is limited by the time required to either reflood the reactor with the high pressure system or the time to depressurize the reactor so that the low pressure systems become effective. Furthermore, the ef fects of CCFL in limiting coolant delivery to the core are not as large at higher reactor pressures. The small break LOCA transient is therefore insensitive to spray cooling and reflooding occurs very rapidly once any one or two of the six low pressure ECCS begin injecting coolant into the reactor vessel. Only medium and large break LOCAs are significantly affected by core spray sparger cracking, and the effect is only significant with the conservative assumption of no CCFL breakdown in the peripheral bundles coupled with an assumed nonuniform spray distribution. 4.4 ANALYSIS RESULTS The most limiting fuel type and exposure combination for the limiting LOCA per the Reload 5 analysis results is a calculated PCT of 1965'F. This is for prepressurized 8x8R fuel at an exposure of 20,000 mwd /t and a MAPLHGR of 12.3 kW/ft. A reanalysis of this limiting case with the unrealistically conservative assump-tions discussed in Section 4.2 results in a calculated PCT of 2075'F. Therefore, a maximum increase in PCT of 110*F bounds the effect of a cracked spray cparger for all fuel types and exposures. A calculation of the maximum PCT for the limiting break with a single failure of a diesel generator using the cracked sparger assumptions of Section 4.2 results in a PCT of less than 1700*F. 4-4

f \ I NEDO-22139 f , l

4.5 CONCLUSION

S A conservative analysis of the effect of one cracked core spray sparger in

   ~ ~

the Peach Bottom BWR results in .a. maximum. increase .in. PCT _of 110*F. - - . . . . Since the Reload 5 analysis shows a minimum margin of 235'F to the 10CFR50 Appendix K limit of 2200*F, the maximum increase in PCT of 110*F can be accommodated with no change in MAPLHGR limit. l l Thus, with cracks in one core spray sparger and with the MAPLHGR limits unchanged, Peach Bottom-2 retains a minimum of 125'F margin to the 2200'F l I PCT limit. This PCT margin is still in excess of the PCT margin taken credit for in the generic study on the ef fect of increased fission gas at higher exposures (References 4-2 and 4-3).

4.6 REFERENCES

l

1. _ _ . ~

4-1 SER, 0.D. Parr (NRC) to G. G'.Shiiwood (dE),' "Re'vidii'6f' G#neral-Electric--- ---- Topical Report NED0-20566, Amendment 3 " June 13, 1978. 4-2 R. E. Engel (GE) to T. A. Ippolito (NRC), " Extension of ECCS Performance l Limits," MFN-077-81, May 6,1981. 4-3 R. E. Engel (GE) to T. A. Ippolito (NRC), " Additional Information 1 ' Regarding Extension of ECCS Performance Limits, MFN-102-81, May 28,1981. l l i l l l l l t 4-5/4-6

                                                   - _.- -              - . _ _ -._. - .- _.- - -                  - ~ - .             _..

4

  .                                            NEDo-22139

• APPENDIX A i STRUCTURAL ANALYSIS OF THE j PEACH BOTTOM 2 CORE SPRAY SPARGER 1

Summary Stress

1. Sparger Pipe (1b/in.2)

                                    - Seismic                                                          853                                 ,

Bending '

                                    - Impingement                                                      698 (No Break)
                                    - Seismic                                                          901                                 l Bending
                                    - Impingement                                                      737                                 l (Break)                                                                                                                  l
                                     - Thermal Mismatch                                               2980 Bending

2. Nozzle 5460 Normal (Weld)

Sh' ear (Ne'id)'

                                         '                  ~ ~
                                                                                                  -- 5700      ---  -            - - -

3. Bracket (Lower) 5140 Normal 1502 Shear 3540 l Normal (Weld) l i

633 Shear (weld) l, 4 Bracket (Middle) 1 9030 Normal 201 Shear  ! 2233 Normal (Weld) 215 Shear (Weld) A-1

.- . . - . . -- ._ - .._ . - . - . . . - _ . - _ _ _ - . - - . _ . . . - _. - . . ~ . NEDO-22139 . A.1 DESIGN LOADS . A.1.1 Impingement Loads (to deflect flow 90*) 4-lN. SCHEDULE 40 PPE y,pg,OV Db. Sc 2 o F,0V D l L 8e [

                                       >                      p = 45.87 lb/ft 0 550*F D"
               ]L j                                                 L2
                      ]L V = 10.0 ft/sec*

F

• 45.87 (10.0)2 (4.5/12)

L 32.2 f=53.4lb/ft=4.45lb/in. A.1.2 Pressure / Flow Loads Maximum Flow = 8000 gpm** (Rated Flow = 6250 gpm) 1 min ft3 Q = 8000 gal / min x 60 sec

• 7.48 gal

               = 17.83 ft 3/sec
    *Very conservative - more realistic value is N2 ft/sec.
   **See page B-5, Appendix B.

A-2

• NEDO-22139 e Maximum pressure in sparger arm AP,,,, = 29 psig @ 6068 gpm

                                                     = 50.4 psig AP ,g = 29 e  Pressure load on sparger segment l

F = AP A A=fd g =f(4.026)2 = 12.73 in.2 l l i F max

                                 =    50.4 (12.73) = 642 lb                                                                        l l

e Maximum nozzle flov

              - .-_ .~ The in.--VNC note-le -has t-he--highest flow._ tate And..will produce the       - - - - - -       .    . ;

maximum nozzle thrust.

                            +          4- a/1s                      n            2         2 A1 = g (1.181 - 0.313 ) = 1.018 in.2 (min.)

w 2 A A; = 7 (1.75 - 0.875 ) = 1.804 in.2 ep l l f o = 62.2 lb/ft 3 0 80*F

1-

                                                                      = 72 gpm @ 6068 gpm test flow 82

• e- - ,

q max 8000 = 95 gpm la al- 7/s q max = 72 -6068

                                 <      >      1.1 s t MIN W,,x       =           = 13.2 lb/sec 6
                               ,         r     14M max   "

13.2(144) = 30,ft/sec @ nozzle exit max " pA 62.2(1.018) A-3

i

                                                                                                    =

NEDO-22139 A.1.3 Nozzle Thrust l HEADER Y Y O l PIPE

    \                                          ()      _x          _    _          _         _z l

Qq Y ) 1 i i ii k--- -g h F > -

                                                                            %     )

e 1 IN, VNC NOZZLE

                              ^

F = LP A + y 8 e , i AP = 25 psi @ 6068 gpm test flow 6P = 25 = 43.5 psi 0 runout 6 86 A=fd , where d = 1.181 in. (the minor dia, of 1-in. straight l internal threads)

                                       = 1.095 in.

A = f (1.181) l l y, max , 13.2(144) = 28 ft/sec @ exit from header pA 62.2(1.095)

                                             )     0%)- = _60, lb F

y

           = 43.5(1.095) + 62.

i l A-4 , 1

                                                                                        -----_---_______________l

l

• NEDO-22139 Fz =" g b +M V = 30 ft/sec (see Section A.1.2) y , 62.2(30) (1.018) = 12.3 lb z 32.2(144)  ;

1 A.1.4 Weight 1 4-in. Schedule 40 pipe W pg p, = 10.8 lb/ft W wa.er r

                               = 5.5 lb/ft W = 10.8 + 5.5 = 16.3 lb/ft l

l ~ -- -- - 10*3 = 1.36 lb/in.- - -- . . . _ _ _ _ A.1.5 Mismatch Due to Thermal Expansion l o* 362.5* ts'

                                                                 \,,..)

l R4' I K4 o 314.1 WRACKET  % ( INLET TEE S SMACKET tSHMOUD CENTERLINE BRACKET g4 0 34.4 CENTERLINE g C ( CS PIPE 27s. { p SMACKET SMACKET

                                .                                                         CENTERLINE             ,

t

                                         ~ CENTERLINE l

l l A-5 l

NEDO-22139 R, = 0+$ 2 = 108.75 in. , R = - = 105.75 in. 2 Shroud = 550*F CS Pipe = 198'F (See page B-2, Appendix B) AT = 352*F ,

                                                   -6 in./in. *F for SST AR = a R AT         a = 9.6 x 10 R = 216 = 103 in, at shroud-to-pipe interf ace 2

For 90' are . . . AR90. = 9.6 x 10-6 (108)(352) = 0.365 in. For segments assume . . . AR = AR 90. (1 - cos0) = 0.365 (1 - cose) l O' I ts* ( INLET TtE (FIXED)

                                                                                                     -sa.4*
                                                                                                     /
                  $3.4'                                   g                                                                      \

f l \ N

                                                                                                                -7s.s*

I I

         . u' -           ( --                                   i A-6 l

l

  ..                                                                                 i NEDO-22139 i

AR IS. = 0.365 (1 - cos 15*) = 0.0124 in. l AR 53.4. = 0.365 (1 - cos 53.4') = 0.1474 in. AR 91.8. = 0.365 (1 - cos 91.8*) = 0.3765 in. AR_3g,4. = 0.365 (1 - cos 38.4') = 0.0790 in. AR

                         -76.8. = 0.365 (1 - cos 76.8*) = 0.2817 in.
                                         /
                                        /
                                         /                                           !
                                         /

w /

                                         /                                           j R

D AR = (20 - sin 20) (cos 20 - 4 cose + 3) 4EIAR g, R (20 - sin 20 - ucos 20 + 4 ucos 0 - 3p) E = 28 x 10 6 lb/in.2 R=Rc= 105.75 in. 0 I = h (4.5 - 4.026') = 7.23 in.' u = 0.2 (coef ficient of friction) 0 y, 4(28 x 10 )(7.23) AR (105.75)3(2e - sin 20 - 0.2 cos 20 + 0.8 cos e - 0.6) A-7

 . - . - . . -         . - _ . - . .              . . - . - - . . .         _ . . . _ . _        - - - . . . ~ . . . . . .-. .. __ _ .. .             . . _ .   =. . .

NEDO-22139 684.7 AR y , (20 - sin 20 - 0.2 cos 20 + 0.8 cose - 0.6) . 684.7(0.0124) = 367 lb y ,., (2nx 180

                                                     -sin 30*-0.2cos30'+0.8cos15'-0.6) 684.7(0.1474) 53 4        -sin 106.8*-0.2cos106.8*+0.8cos53.4*-0.6f 53*4'*(2nx                    180
                                    = 120 lb 684.7(0.3765) g 91'8 - sin 183.6* - 0.2 cos 183.6* + 0.8 cos 91.8' - 0.6')

91'8',(2nx 18O

                                     = 91. Ib 684.7(0.0790)                                                 _,

y ,

                                           '(2nx38.'                -sin 76.8'-0.2cos76.8*+0.8cos38.4'-0.6) 0                                                                                                              ,
                                        = 155 lb 684.7(0.2817) y-76'8' ,_/                       76 8        - sin 153.6* - 0.2 cos 153.6* + 0.8 cos 76.8' - 0.6')

(2n x 80  ! i

                                          = 96,Ib                                                                                                                         <

Flow-Induced Vibration - Natural Frequency f A.1.6 The vortex shedding frequency, fn, is given by: i fy D . y

                                   = 0.21 i

A-8 , w - r w --- ---

       .             -. -                           -- -. . _ .            . - . _   ., . - _ - . . - - _ - - - . - . - . - . - -                  . ~ . - . . _.-

i

• NEDO-22139  !

V.= velocity past the shroud wall = 10 ft/sec* D = sparger pipe diameter = 2 It j f

                          ,0.21(10) = 5.6 Hz v        4.5/12 General Electric design basis requires natural frequency:

l fn 13 v . Assume Calculate the natural frequency of the unsupported sparger segment. the segment acts as a cantilever and has a uniform load, w (1bs per unit length):

                          /                 w g
                          /
                          /

kifif kif if f n wL

                          /

' ~ ~ ~

                          /                                                                                                       . . . . .                 . , , _ ,
             ~
                          /

C '  ? Kn = 3.52 2 6 I = 7.23 in.' g = 32.2 ft/sec E = 25.75 x 10 lb/in. L = E x w x 105.75 = 28 in. (distance from crack to nearest support 180 bracket)  ! w = 1.36 lb/in. (Section A.1.4) 6 g , 3. 5 2_ 25.75 x 10 (7.23)(32.2)(12) = 167 Hz - n 2n 1.36(28)' Ratio = = >3 v

                   *Very conservative - more realistic value is s2 ft/sec.

A-9

' NEDO-22139 Calculate the natural frequency of the sparger by examining the longest seg-ment between support brackets. Assume this section has'a uniform load, w, and I both ends are simply supported:

• K

                                                      !        tn. E.2n       S 9fifi     fifif 1f if if 1 7 O 1          [

7,4 h h Kn " 9*07 I = 7.23 in.' 7 6 E = 25.75 x 10 lb/in.2 8 = 32.2 ft/sec 8 L= y80 x u x 105.75 = 71 in. w = 1.36 lb/in. 6 9.87 25.75 x 10 (7.23)(32.2)(12)-= 72,Hz f,= 2n 1.36(71)4 Ratio = = {>3 v Calculate the natural frequency of the sparger by examining the longest seg-ment between support brackets ignoring an intermediate support. This case is the same as the above case except that L = 2 x 71 = 142 inches. f = = 18 Hz n i

                                       =

Ratio = 5.6 > 3 i i V l 1 a.- 10 l

           ._.m  . _ _ . .         _ . _ . - _        _ _ . _ . . . . _ _ _ . _ - _ _ _ _ _ . _ _ _ . - _ .                         .-._ . _ . - .._ .._ _ . _   _.

i' i NEDO-22139 Calculate the natural frequency of the sparger by considering the sparger arm l as a " free-free" beam (or floating ship). Assume the arm has a uniform load, i

e, and is free to rotate at the three support brackets as shown below

! K

%- g n

2w n Ella.

                                                                                                                        "I'4

? / l

                                                ,/
                                                                                                                     = 61.7 Kn"         2
                                 /                                                                                            6
                               /                                                                        E = 25.75 x 10 lb/in.2 I

I = 7.23 in.0 g = 32.2 ft/sec A

                               \

w = 1.36 lb/in. (\ L = '18O x x x 105.75 = 180 in. 6 l f

                                 ,61.7             25.75 x 10 (7.23)(32.2)(12) = 70 Hz                          -

n 2n 1.36(180)4 I s, Ratio = = 5.6 >3 v A.2 STRESSES DURING NORMAL OPERATION AND DURING CORE SPRAY INJECTION A.2.1 Sparger Pipe A.2.1.1 Impingement Load and Seismic Impingement only w g = -4.45 lb/in. (upward) (Section A.1.1) A-11 P

NEDO-22139 . Seismic Only - Assume 3g (Very_ Conservative) , w, = w 1 3w w = 1.36 lb/in. (Section A.1.4) lb/in. (upward) w, = 1.36 - 3(1.36) = -2.72 w, = 1.36 + 3(1.36) = 5.44 lb/in. (downward) Impingement + Seismic w = -4.45 - 2.72 = -7.17 lb/in. (upward) T w = -4.45 + 5.44 = 0.99 lb/in. (downward) Assume No Break j For simplicity, assume continuous beam - three equal spans.

                                                            "                                                              l I
                 ,r u'r 'I v'r 1r u 'r it 'r 'r 'r 'r 'r f 'r 'r "                                                         l AL AL                           AL 4L
                                      > c                     t       ; c                t        7 4          i Rg = M we                R, = 1.10 we                  RC = 1.10 we                 RD=0M OA0 wt                                              ;

0.40 we 0.30 we SHE.AR tbl

                                     %s 0.60 we                      0.40 we 0.80 we
                                        -4.10 wt 2                    -0.10 wt 2
                                           /\ \                         /
 -.m...,                              /                         '
                                                                  /

N\

                                  /

7 M+0.026 we % N  % %_,/,

                                                                                                  /
                                                                                                    /

N%_

                           . W8                                                          +0m J 2 A-12 l

e-

  '                                                    NEDO-22139 i = 3880' x w x 105.75 = 71 in.

M ,x = 0.10(5.44)(71)2 = 2742 in.-lb

                                                                            = 2.25 in, o=           I = 7.23 in.4                       c=
                     =                   = 853 lb/in.             (Seismic) o 7f2
                                  '       = 698 Ib/in.2 (Impingement) e,,x = 853 54 Assume Break Assume two equals spans, uniformly loaded with end moment M3 and force P 3 at the third support.

P 3 w

           '               I                      f i          If1      1r t 1f 1r1        fl         if fif                           _,

M 3 Y u,=

                                      ; ;                   e        p.+t ,3 4         e R                                R A                               2                               3 1

From the theorem of three moments . 2 3 M117 L 1 1 2 H 32 "

                                                                  "11           +
                                                                                   "22 1

1 2 T+T+ 1 2 I 2 4I 1 4I 2 M y =0 1 7

                                       =1 2        Il"12                "1 "

• 2 A-13

6 NEDO-22139 . 4M 1 M1 3 gg 3 2

                   +

• 2I I I

                          ~

2" , M is caused by the cantilevered section of pipe between the support 3 bracket and the break: 2 wt 3 N

• 3 2 h.

1.ikewise, P is caused by the cantilevered section: 3 P *"L 3 3 For Seismic . . t = 71 in. w = 5.44 1b/in. 13 = 28 in. P3 = 5.44(28) = 152 lb M ) = 2132 in.-ib M3- 2 M2 5 44po' - 2132 . 2 95 in.-1b c..x)

               ..y              1    7.23 in.4       c.4;5- = 2.25 in.

o 2895(2.25)- = 901_ lb/in.2 max 7.23 i d A-14 i . _

NEDO-22139 For Impingement . .. I w =,l-4.45l=4.45lb/in. P3 = 152 l= 125 lb 3

                  = 2132
                             ,44
                                    = 1744.in.-lb M2 = 2895 5.44
                                    = 2368 in.-lb o,,    = 2368 3
                                    - = 737 lb/in.2 Determine reaction loads for seismic + impingement a = 7.17 lb/in.
                                 +     -   = 201 +
                                                       .17(

2

                                                            }+       ~

1

                                                                             = 441 lb R3=P3 + "2
                   = ."   _      =  7.17(71) , 38 5 = 201 lb I

3 2811

                          +2         -

(M' = M2 ) = 7.17(71) + 2 1

                                                                                = 577_lb R2=wt R +R2+R3 = 1219 lb checks u (21 + 13) = 1219 lb A.2.1.2 Differential Pressure                                                        I AP,,    = 50.^ psi l

R,= 4.5/2 = 2.25 in. A-15

NEDO-22139 . l Rg = 4.026/2 = 2.013 in. .

                                                                                                                        )

t = 0.237 nom. c.1,. = 0.237 - 210.003 (cerrosion a11 ance>> = 0.231 in. e Hoop Stress a= = 50.4(2.013)- = 440 lb/in. 1 e Axia1 Stress , o= " 2

                                         =       220 lb/in.

t A.2.1.3 Mismatch Due to Thermal Expansion' M = WR sine - UWR (1 - cos0)

                                                  /
                                                  /

j u = WR [ sine - p(1 - cos0)]

                                                  /
                                                  /

R W

• I f

mw Assume p = 0.2 (Coef ficient of f riction) R = 105.75 in. See Section A.1.5 for loads at each bracket: A-16

l ' NEDO-22139 l e 15' Bracket M

15. = 360(105.75) [ sin 15' - 0.2(1 - cos 15*)]

I l = 9590 in.-lb (Maximum) e 53.4* Bracket T M 53.4

                                  = 111(105.75) (sin 53.4* - 0.2(1 - cos 53.4*)]

l

                                  = 8580 in.-lb 3

o 91.8* Bracket 4 I l M ) 91.8. = 79(105.75) (sin 91.8* - 0.2(1 - cos 91.8*)] \ I

                                    = 6630 in.-lb                                         ?

I . e -38.4* Bracket M

                             -38.4. = 147(105.75) [ sin 38.4 - 0.2(1 - cos 38.4*)]
                                     = 8980 in.-lb

- e -76.8* Bracket ^ M

                              -76.8   - 86(105.75) (sin 76.8* - 0.2(1 - cos 76.8*)]        l
                                      = 7450 in.-lb l

i o . Ibi c= 2

                                                  = 2.25 in.

max I i i ) i I = 7.23 in.4 I i I I

                ." max . 9590(2.25) 7.23           = 2980 lb/in.
 )

l I A-17

                                                               ~
                                                                                        , I

NEDO-22139 , 1 A.2.2 Nozzles A.2.2.1 Nozzle Thrust Y Y Y

           )                                 )

l , f' l l

                                                                                     -z                         l h                 A x      -            -                                       '
              , .n +              +

t.ss L

                          '              b
       \

0.12 ( p 1 f e-J f V ARIES - Y ASSUME 15* 3.29 - 4 15,

                              ,m          r k' eld properties:

I=h(1.76 0

                          - 1.524 ) = 0.209 in.'

4 4 K = 32 1 (1.76 - 1.52 ) = 0.418 in.' 2 A = { (1.76 - 1.52 ) = 0.618 in.2 l 1 1 76 = 0.88 in. t = 0.12 in. r= = 0.88 in. c= 2 2 Loads: F = 60 lb f ' = 12.3 lb (See Section A.1.3)  ! y z P = 43.5 psi A-18

,.o

   ~'

NEDO-22139 The resulting loads at the weld are . .. F =F = 60 lb axial y F shear

                       =

F,' = 12. 3 lb T torsion

                          = 3. 9   F,' = 3.29(12.3) = 40.5 in.-lb H

moment

                         = 1.96 Fz ' = 1.96(12.3) = 24.1 in.-lb The stresses are conservatively calculated as .              . .
                       "mc             r                       60 (0. 8 8) , 0. 618 4 4.35(0.88)
             ,y ,.- I        , A , F_r, 2t ,
                                            - 24.1 0.209                      2(0.12)
                  =   101 + 97 + 160 lb/in.2 o    = 358 lb/in.2 , 156 lb/in.2 y

Tc F T xy

                    =
                             +af               a = 2 (thin wall cyl.)
                    , 40.5(0.88) + 2 12.3 = 85 + 40 0.418           0.618 T

xy = 125 lb/in.2 A-19

NED0-33139 A.2.2.2 Differential Pressure . Assume 360' break, nozzle loaded by bracket, 3 f, WCTION 1.21 P = M2 LS

                                                }

' { I

                             ~

f3 k k LJ

                 '                                                              t.se

{

     \

k

   ..n                          J+     /
                                   /                                       2.88 The resulting loads at the weld are .        . .

F = F = 642 lb shear . T = 2.68F = 2.68(642) = 1720 in.-lb torsion M ,g = 1.96F = 1.96(642) = 1260 in.-lb The stresses are conservatively calculated as . . . Mc m ,P2 , , 1260(0.88)_ 43.8(0.88) = 5300 + 160

       ,y , , 1          2t       0.209        2(0.12) o = -5,140 lb/in.       , 5,460 lb/in.2 y

A-20

• NEDO-22139 i I l

T F ( a = 2.0 ' T 3, = - [c +of l l

                         , 1720(0.88) + 2 642 = 3,620 + 2,080 0.418             0.618 n

t = 5,700 lb/in.' xy A.2.3 Bracket (Lower)

                                               ?/

O '  ; (__ 2.25 s f G A_ y T R

                                                                                                ,,   o lf     R                                                   Z_I X

6 l

                                                                                         ~

() Ji , t' I L e

                    \                                                                                       l h

[ . .

                                        '                     LOWER BRACKET 1 = 3.26 in                   L = (1 + cos 45*)(2.25) = 3.84 in.

l

f. ' =

2

                                      + (1 - sin 45*)(2.25) = 2.29 in.

l b = 0.38 in h = 0.25 in. l A-21

NEDO-22139 4 A.2.3.1 Seismic and Impingement  ! Ry = 577-lb R, = 0 R, = 0 (Section A.2.1.1) , (Conservatisms - Uses highest bracket load at weakest bracket and assumes ceismic and impingement downward.) - b Maximum stresses in the fillet veld . . . Avg " 0. 3 26) " b .

                                  , 36 M ,                y , 3 6 (3.84)(577) = 3540 lb/in.                                                 j i
           , Bending                        2 ht               0.25(3.26)2 h1                                                                                                 '1 A.2.3.2 Mismatch Due to Thermal Expansion                                                                                             ;

R = 367 lb R =pR x

                                                                    = 0.2(367) = 73.4 lb X                                        2 (Section A.1.5)

R =0 y Maximum shear stress in the fillet veld is . . , i R g g

             ""I ht  6      x     +      R,   + T  (b  +  h)(1     - h)h where M = L' R,                                                                                                                     I 2.29(73.4) 6 (367 + 73.4) + 26 (0.38 + 0.25)(3.26                                  - 0.25)(0.25)
              " " T 0.25(3.26)                                                                                                               i i

A-22 t

                                ,                               --    ,      -y _ . .__,.. __

8 " NEDo-22139 ( , w = 382 + 251 = _633,ib/in.2 Maximum normal stress in weld is . .. 2 2L 2 ,(b + h)2 , 36 M g max . $2 bhi + h1(b + h) 2 h1 2 where M = t' R x

                          ,, d        (367) max      2     0.25(3.26) 0.25(3.26)2
                                                        + 36 (2.29)(367) 73.4                   2(3.84)2 + (0.38 + 0.25)2 2
                             + 0.25(3.26)(0.38 + 0.25) o ,x = 318 + 1342 + 779 - 2439 lb/in.2 Maximum normal stress in the plate is .                . .

xy + LR, C,x R x E IxC max " T + I xy I zx A = 0.38(3.26) = 1.24 in.2 bt = ) = 1.097 in.4 I xy =2 C xy "

                                ~

2

                                       " 1*03    1"'

7

                           , tb , 3.26(0.38) = 0.01491 in.

l zx 12 12 A-23

                                                 . = _ . _ . . _ _ . . _ . . . . . . _ . -                 - . . . _ . _ = _ . _ _ _ _ . _ _ . _ . . _ _ . . . _ _ _ _ _ _ _ .
                                                                                                                                                                                   '.., i J                                                                                           NEDO-22139 4
                              =

0.38 = 0.19 . C 1 sx 2 .

                        , max , 367  1.24_ , 2.29(367)(1.63)     1.097                       , 3.84(73.4)(0.19) 0.01491               = 300 + 1250 + 3590 k

i 4 L o max

                                 = 5140 lb/in.2                                                                                                                                         ,

Maximum shear stress in plate is . . . s R x + R, + L ' R, ( 31 + 1. 8 b) 367 + 73.4 T " bi g23 2

                                                                                              " 0.38(3.26)
                                , 2.29(73.4)(3 x 3.26 + 1.8 x 0.38)                                                                                                                     '

(3.26)2(0.38) T = 356 + 1146 = g lb/in.2 6 I i 4 i A-24

l , [ l

     .                                                     NEDO-22139 A.2.4  Bracket (Middle) l 1

1/ se* 2.3 l [ 3X E _n ~, T N' ' V )

                                                      -        A n                                  /

t'

                      'F                              /                                  "      p If                                 /                                 '

S /

                                                       /

b ' db < >

                                             \/,-/

i ( p

                                                        /
                                                        /      V . . ._ . . .

f' [ .

                                                        /
                                                        /

g, f R'2

                                                        /
                                                                                    &    &6
                              %            l             /

a J

                              -                     -    /

L L = 12.12 - 2(2.25) = 7.62 in, b = 0.38 in.

                          ~
                                + (1 - sin 30')(2.25) = 4.94 in,                 b = 0.25 in.

L' = 2 I L = (1 + cos 30*)(2.25) = 4.20 in. 1 = 7 (2.25 + 1.96) cos 15' + (1.50 + 1.18) sin 15' - 2.25 - 2.51 in. A-25

_ . . . .. _ . . _ . . . . . . _ _ . . - . - . . _ . _ . . _ _ . - . - _ _ . _ . _ _ _ . _ _ .m _. _ __ _ __ _ . . . ___ NEDO-22139

                            'L p  = 2.25 - (2.25 + 1.96) sin 15' + 1.50 +                                           cos 15'                                                    ,

Lp = 3.18 in.  : I A.2.4.1 Pressure Load Only F = 642 lb (Section A.1.2) Shear Stress'(Neglect torsion - small) N " = 238 lb/in. (Weld) Avg " 0. 62) T = 222 lb/in.2 (Bracket) Avg " " 0.38 .62) d Stress Due to Bending F 2 + (b'+ 2 h)2 max " ht(b + h) F 642 2(3.18)2 + (0.38 +2 0.25)

                                     " 0.25(7.62)(0.38 + 0.25) o    = 2420 lb/in.2 (Weld)

0. = 0.19 in.

c,,x = c= 2  : t 1 = 4b' = 7e2g.>e>> = e.e>4e4 1. 4 1, M = (LF - h) F . 1 A-26

,. . . . . .- . - - . . .- . - . - _ . - - - _ - . - . ~ . - . - . - . - . - . - . . . . . . . - . - - . . - . . - - . . - . . - t . l , NEDO-22139 i

            .             o max " (3.180.03484                          - 0.25)(642)(0.19) = 10.260 lb/in.2 (Bracket) r A.2.4.2 . Mismatch Due to Thermal Expansion Only l

t i Assume: R, =R x = 155 lb (Section A.1.5) i l. R =R = 0.2(155) = 31 lb

                                *1,             *2 Shear Stress
                                                                   +R d(R      *1 2         6          62 "avs                  2            ht                         2 (0.25)(7.62) w
                                   ,yg
                                           = 2j! lb/in.                      (Weld) l
                                                           +R.

(R*1  !

                                                                      *2                   62 l

• avg " bt ~ " 0.38(7.62)~

l n,yg = 2j( lb/in.2 (Bracket) Normal Stress  ! l R +R R +R g

                                           /I         *1            *2                *1     *2 2L   2 ,(b + h) 2           ht                        ht(b + h)                           2 i

i /I 310 61 2(4.20)2 + (0.38'+ 2 0.25)2

                                   # " "i" 0.25(7.62)

• 0.25(7.62)(0.38 + 0.25) x i

                                      = 115               302 = 445 lb/in.2 , -187 lb/in.2 (Weld) l                                                                                                         A-27 l

l l

_ _ . - . _ . - . . _ _ _ . . _ . _ . _ . . _ _ . _ ___ __.___._._ _.-._-._ _._..__ _ _...-..,_.__~._. _.__

                                                                                                                                   -NEDO-22139                       ,   ,

R x

                                                               +R x             (L'- h) It #                               -R c

• 1 1 2 bi 2* I c = 0.19 in. I = 0.0348 in.'

, 310

                                  , , 0.38(7.62) g (4.20 - 0.25)(62)(0.19)                                     0.0348                                  = 1101 1340 ,

o = 1450 lb/in.2 , -1230 lb/in.2 (Bracket) A.2.4.3 Combined Stresses During CS Injection , Shear Stress n Avg

                                                  = 238 - 23 = 215 lb/in.2 (Weld)                                                                                              ,

i w Avg

                                                   = 222 - 21 = 201 lb/in.2 (Bracket)

Normal Stress e = 2420 - 187 = 2233 lb/in. (Weld) l o = 10,260 - 1230 = 9030 lb/in. (Bracket)  ; 1 1 A.3 REFERENCES  ? GE Drawing 731E779, " Core Spray Sparger."  ; i GE Drawing 761E506, " Core Spray Sparger." l l Roark, R. J. , " Formulas for Stress and Strain," McGraw Hill, Fourth Edition, i 1965. A-28 l l

                                                                                              . . - - . , . . , - , - - - . . _ _ - - - - - _ .                  ~

     +                                                                                             NEDO-22139 t

Hopkins, R. B., " Design Analysis of Shafts and Beams," McGraw Hill.

                  " Machinery's Handbook," The Industrial Press,16th Edition,1959.
                               -          -                    - - - - - -    - - - -                                         ~ . . -        . _.                 ._,__ __ , . . . _                     ._

Blevins, R. D., " Flow-Induced Vibration," Van Nostrand Reinhold Co., 1977. Shields, C. M., Wade, G. E., " Core Spray Distribution No. 17, 251 Standard Plant," NEDE-13006-4, December 1, 1970. l l l l I l 1 l l l i 1 1 l l l l l l 1 l l A-29/A-30

                    -                               -                                 , - - - - -               , -                  -  -e .      ,                              - -

-**' ^*--- --., u _ ____. 4 _ _ _ 0 e

                                                % 4 e

b 4 1 1 i l f

NEDO-22139 APPENDIX B SPARGER TEMPERATURE CALCULATIONS B.1 SPARGER TEMPERATURE Heat transfer coefficients for inadvertent spray injection are from pages B-3 and B-4. 4 -De  ? hg = 5037 Btu /hr - ft - *F - r, 7, h,= 365 Btu /hr - ft 2 , .7 h i / K = 10 Btu /hr - ft *F n O (304 sst @ 200*F) P t = 0.237 in. D g = 4.026 in. D, = 4.5 in. Tg = vater in sparger = 80*F T, = vater outside = 550*F Ag =n 1 = 1.054 ft (1 ft long section) 2 A, = n g'j 1 = 1.178 ft (1 ft long section) A =n 4.0 6 + 4.5 1 = 1.116 ft2 (1 ft long section) The themal resistance, R, is: 1 l_ . R 1 + 1 + t

                              ^ oo     ^p B-1

l

                                                                                                                                      . g. i I
                                                                                                                                             \

) NEDO-22139 , Q= To_Tg 1 1 t

                 'A h 4    Ah ,AK                                                                                                     ,

ii oo p Ah ho ~ 1) ATfilm outside " O Ah 1 1 + 1 Ah gg Ah AK p oo

                                                                     ~

Ah o 1) 0 film inside " 1 1 t Ah +Ah +AK p

            *"Ah +Ah +AK 1                  +         1                                 0.237/12
             * "'1.054 (5037)                          1.178 (365) + 1.116 (10) x = 0.000188 + 0.002326 + 0.001770 = 0.004284 0.002326 (550 - 80) = 255.F                          .

ATfilm outside " 0.004284 i Outside metal temperature = 550 - 255 = 295'F r AT = 0.000188 (550 - 80) = 21' film inside 0.004284 Inside metal temperature = 80 + 21 - 101*F 01, Average sparger (pipe) temperature = f AT = 550 - 198 = 352*F_ Bracket to pipe B-2

       . --        --      .-             -        ...                   .. -                     _ =-

 .                                                                           _ = . -    ..       . . .

• 1 l

I NEDO-22139 l ( , 4 In practice, the core spray pumping system cannot inject into the reactor until the pressure reaches 300 psia, where T,,g = 417'F. In this case, the AT Bracket to Pipe is less than 337'F (417 - 80)*. Thus, the above calcula-l tion bounds the inadvertent injection case. It also bounts the case of core spray operation during LOCA for the same reason. B.2 CONSERVATISMS

1. Bounding for reason described above. j l

2. Assumes steady-state conditions (Q =Q = Qg ).

! 3. Neglects heat conduction from pipe. 1

4. Assumes runout flow.

l I l B.3 REFERENCES l l l

1. Kreich, " Principles of Heat Transfer", International,1969.

2. Welty, et.al., " Fundamentals of Momentum, Heat and Mass Transfer",

John Wiley, 1969. l B.4 HEAT TRANSFER COEFFICIENTS B.4.1 Inside Sparger Arm (Near T-Box on Leng Side) l Assume average film temperature = 90*F D, = 4.026/12 ft A gy,

                                                       =f          = 0.0884 ft o = 62.1 lb/ft 30 90*F I

I *6T film inside and AT pg p, are ignored. l l B-3 1 l l . - - _

, s. NEDO-22139 2 v = 0.833 (10-5) ft /sec . l W = 7980 gpm = 1102 lb/sec Total l W Arm = 1102 360 (97*5 = 298.5 lb/sec Y" = 54.4 ft/sec l

                         ~~ " 0.088                    62.1) i Nu = 0.023 R e                           P r

l l R =E0 e y ! 4.026 l - 54.4 ) 6 R = I e 12 (0.833(10-5f,,=2.19x10 l P7 = 5.20 0 90*F N = 0.023 2.19 x 10 0 (5.20)1/3 = 4,707 9 l I D = 4.020 gg Id! K

                         . Np K = 0.359 Bru/hr - ft                         "F G 90*F F
                                '                     $9)      = SfL37,Beu/hr - ft                                - *F hy      =

4 026 1 l I ! B.4.2 Outside of Sparger Arm I

1. Assume average water velocity is N2 ft/sec.

i J

2. Assume average film temperature = 420*F.

i i ' 3. Assume that heat transfer is like a cylinder in cross flow. B-4 I l 3 , m,.-,,c,e-- -- . _ , _ . ._ - ~

    .        . . _ _ .      .~.__._.m.            _ . . - . _ . _ . _ _ . _ _ . _ , _ . . . _ _ _ _ . . _ _ _ _ . . . _ _ _ _ . _ _ - . . _ _ _ _ _ _ _ _ _

J f . *

                                                                                                                                                                            'i f                                                                                                                                                                           !

NEDO-22139 , 1 *

                                                      ~

I

• v = 0.169 x 10 ft /sec @ 420'F l

i K = 0.375 Btu /ft - hr 'F i 1 2 i Pr = 0.932 ] i ) 4 C

• P

                                                                                                                                           .31 j                                 K
                                         =

0.35+0.56kR) e . I 1 1

                                      =

4.5 ft 1 D o 12 (

• 5 R =

I i

• 12(0.169x10-5j=4.438x10 u "

0.9320 31 5 h, = 0.35 + 0.56 4.438 x 10 .

                                           .5/ 2 1

2 J' ho = 365 Btu /hr - ft - *F f. I 1' l j B.5 PUMP HEAD / RUNOUT l 1 i l 1 I Shutoff Head = 300 psia (Q = 0) , s 8

                                             = 6250 gpm @ 125 psia QRated 2

P=P SH

                                                -CQ I

where P SH = shutoff head 125 = 300 - C (6250) B-5

                                                                                                                                                                             )

                                                            . t, ,

NED0-12239

          ~l$    =4.48(10~6) psi /gpm C=          2 6250
 @ P = 14.7 psia (Runout)
                     ~

SH , 300 - 14,7 = 7980 gpm 1 8000 gpm q . Runout C 4.48(10-6) y = 7980 (62) = 1102 lb/sec Runout 60 (7.48) 5 B-6

NED0-22139 APPENDIX C FLOW VELOCITY CALCULATIONS This appendix describes the calculations for the flow velocities supporting statements in Section 3.4.2.1 of the text. C.1 FLOW VELOCITY IN BYPASS REGION Assumptions:

1. The plant is operating at rated power (3293 MWt) and flow 102.5 x 106 lb/hr.

2. The flow in the bypass regions is homogeneous.

6

3. The bypass flow fraction is 12% (12.3 x 10 lb/hr).

4. The water in the bypass regions is saturated.

5. There is no down flow in the bypass region. This assumption is dis-cussed later.

There are two parallel flow paths in the bypass region--one is between 'the fuel channels, and the other is between the core shroud and the outermost fuel assemblies. The flow areas for these paths are shown schematically in Fig-ure C-1. The simple analysis that follows will give an estimate of the rela-tive flow velocity in the neighborhood of the spray sparger. The flow Path 1 is between the core shroud and the outermost fuel channels. area along path 1 changes from Ay , between the bottom and top of the active immediately above the top guide: fuel, to AS at the top guide to A6 I"' Ag _ = 5261 in. .A5 = 3720 in. , A6 " 91" C-1

                                                                                              ~

l. NEDO-22139 I The flow area along path 2 changes < Path 2 is between the fuel channels. i A at the top guide to A4 above the fuel channels: . from A2* 3 A 3 = 2028 in.2, A 4 - 27504 in. 2 = 3918 in. .A From the geometry and the flow areas, K for path 1 is approximately 0.3 and K for path 2 is approximately 1.0: (KWy y )/A5

                          "     N IA 3 22 (0.3)W 1         (1.0)W 2 3720            2028 Wy = 2.5 W 2 0.40 W y     =W 2 6

1.40 Wy = 12.3 x 10 6 Wy = 8.8 x 10 lb/hr The velocity in the bypass region between the core spray sparger and the fuel assemblies is then: 6 V =.W y /0A 5 = 8.8 x 10 /[3600 x 45.8 x (3720/144)] = 2.1 ft/see The fluid in this region is primarily saturated liquid. The fluid velocity in the periphery of the core bypass region was conserva-In actuality, there probably is downflow in tively estimated at 2.1 ft/sec. this region. The total pressure drop across the top guide is predominantly In some portions near the top of the core bypass due to the elevation head. - Because region, boiling may occur, reducing the elevation pressure drop. C-2

s

• i I

NEDO-22139 i n.

         .)                                                          '

I' there are no heat sourcis'in the non-fueled peripheral regions of the core j bypass, boiling would not be expected in the vicinity of the shroud. Thus, ! some downflow or crossflow in the peripheral regions toward the central region 3 would be anticipated to balance the density differences. 4 C.2 FLOW VELOCITY AT TOP SURFACE OF CORE PLATE l d i J V = (WTotal Bypass)/pA 6 Since WTotal Bypass = 12.3 x 10 lb/hr A = w/4 (D - Nd)2 l- D = inside diameter of shroud - 204 in. l N = number of control rod guide tubes = 185

• d = outside diameter of control rod guide tube = 10.875 in.

o = density = 45.8 lb/ft 3 4 i Then i 6 2 V = (12.3 x 10 )/(3600 x 45.8 x w (204 - 185(10.875)2)f(4 x 144)) l $ V = 0.69 ft/sec i C.3 FLOW VELOCITY AT THE TOP OF THE FUEL ASSEMBLY HANDLES i 6 0 6 W Total

                                          = 102.5 x 10               - 12.3 x 10 = 90.2 x 10 lb/hr i

< A'= na i n = number of fuel assemblies = 764 ) if a = area associated with each fuel assembly = (6) = 36 in.2 1 i C-3 l

  . . .    . . . . - - . .         . - . . -         _ . - . . ~ . . . .       .  . . .      , . - . - - . . .      - - . .         . ..-                 ~ .. . . . . - . .
                                                                                                                                                                  ,y*

• NEDo-22139

;                                                                                                                                                                      a The equivalent single phase velocity is:

V = (WTotal)/pA , 1 Then 1 0 V = (90.2 x 10 )/(3600 x 45.8 (764 x 36/144)) = 2.86 ft/sec At this location, the fluid is a mixture of steam and water. Therefore, to calculate the lifting force due to the mixture, a two-phase friction multiplier must be used:

                         $ m = 1 + x (p f /o - 1) mass flow rate of steam x = quality = total mass flow rate 6                     6
                              =  (13.4 x 10 )/(102.5 x 10 ) = 0.131 3

o f = 45.8 lb/ft e o = 2.35 lb/ft 3

                           $ = 1 + 0.131(45.8/ 2.35 - 1) = 3.42

• The total lifting force on a section of core spray pipe per unit length is:

F=C AO D f *2 V'I(2 ) 8 4 C-4 _ . _ . _ _ ~ _ . _ , _

i, . ) ,. NEDO-22139 a where

    .g CD = drag coefficient = 1.2 2

A = area = (4.5 in. x 1(ft/ft))/12 (in./ft) = 0.375 ft ffg Then: F = 1.2 x 0.375 x 45.8 x 3.42 x (2.86)2 / (2 x 32.2)

• 9.0 lb/ft I

l l l 1 l l I C-5

                                                                                                                                              .' f, k

• *e ..

! i 4

l NEDO-22139 l 4 4 i

                                                                                                                                                         =        j 4           ,

l 1

• 1 i I l

j l l i, l 5 1 1 1

                                                                                                            ~

i A #4 q 6 1 4 I I 1 i

                               -->-         A 5
                                                                                              .         =        3=                                               ,

l _ i 1

                                      ', PATH t                                                         'P ATM 2 1

l 1 1 1 1 l-

                                   +         A                                                   <             ^2*                                                '

_3 D- - J Figure C-1. Flow Paths s e C-6

'e '

PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P C DOX 0000 PHIL ADELPHIA, PA,19101

             ]>,E,ya ,.(('

ovember 8, 1985 Docket No. 50-278 Mr. John F. Stolz, Chief Operating Reactors Dranch #4 l Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555  ! l i

Dear Mr. Stolz:

I.E. required, in part, Bulletin that80-13, " Cracking in Core Spray Spargers", examination of the core spray in the event cracks are identified during l be submitted to the Of fice of Nuclear sparger system, an evaluation shall Reactor Regulation for review and Licensee Event approval prior to return to operation. As reported in Report 3-85-14, performed during the current Peach Bottom Unitinvessel inservice inspection, 3 refueling outage, identified heat af fected zonetwo cracks of the 'A' on the piping to junction box weld reactor vessel. core spray piping just inside the Unit Attachment I titled, " Peach Bottom Atomic Power Station 3 Core Spray to ca f ety evaluation Line Cracking Safety Evaluation", provides a support Peach Bott om Unit 3. restart and continued operation of Although the evaluation concludes that no modifications are required in order to assure safe operation, Philadelphia Elect ric Company has installed two brackets at the location of the cracks operability and safeto provide f urther assurance of core spray sparger reactor operation. been installed on the '~ Similar brackets have wore identi fied in that core iping. spray header although no cracks D: comber, The 1985;refueling outage is scheduled for completion in thoroforo, would be appreciated. prompt consideration of these mattere 051MO20 051100 R ADOCK 05000270 PDR hd>\ {\

                                                                                             \\

e = Mr. John F. Stolz November 8, 1985 Page 2 not hesitate If we tocan provide contact us. any additional information, please do Very truly yours,

                                                                                      )
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( cc Dr. T. T. P. E. Murley, Admini s t ra t or, Region I, USNRC Johnson, Resident Site Inpsector

it' h HDE-216-1085 DRF-E21-00090 October 1985 PEACH BOTT0H ATOMIC POWER STATION UNIT 3 CORE SPRAY LINE CRACKING SAFETY EVALUATION Prepared by: W P. T. Tran, Engineer Application Analysis Services - Verified by s C. H. Stoll, Principal Engineer Application Analysis Services Reviewed by: h R. R. Blooderand, Senior Licensing Engineer Licensing Services Approved by: ' G. L. Sozj:i, M6hs'ger Application Anal,ysis Services 051113CK M05000270 B PDR C ' G E N E R A L $, ELECTRIC NUOfA CNERGY 8USINESS OPEAATIONS 1 GCNERAL f tECTUC COMPANY e 17) CygTNgg AgNyg a gm ggg, gjyroggg4 939g3

a c - O t PEACH BOTTOM ATOMIC POWER STATION UNIT 3 CORE SPRAY LINE CRACKING

                              !                                    SAFETY EVALUATION September 1985 During the current refueling and maintenance oucage, invessel inservice

• inspection located a crack on one side of the piping to junction box weld heet affected zone of the 'A' core spray Line (header) at Peach Bottom Unit 3.

Bulletit 80-13.The crack was identified during inspection in accordance to IE were performed to Subsequent verify that to identifying the crack air bubble tests the test verified that the crack was through-wall crack was through-wall. The air bubble and also revealed another through-vall 1). crack on the opposite pipe to junction box weld (see Figure Both cracks are in the piping in the weld heat affected zone (RAZ). One crack appears to run approximately 180 degrees in length and appears to be approximately 120 degrees through-wall. The other crack appears to be about appears to120 degrees in length but the through wall air bubble leakage i resemble a pin-hole. 1 l General significance Elcetric has performed of these cracks. evaluations regarding the safety { follows: A summary of these results is provided as

A. Area of discussion a

1. Analysin significance. has been directed toward the following areas of n.

b. An estimate of the leakage flow through the cracks

Structural integrity c.

d. Emergency core cooling system performance limits i j

Structural repairs {

e. Loose parts I

B. Renults i S i 1.a. Crack Leakage Estimate i i From test, a review of both inspections, the visual and air bubble  ! the estimated leakage through both cracks appears to be 4 less than half the leakage flow through the 1/4-inch vent hole present in the tee-box. 1 A bounding estimate indicates that the  ! flow through this core spray operation. vent hole would be less than 13 gpm during i During normal reactor operation the flow through this line is expected to be negligible. Therefore, during the core spray injection phase of a*LOCA the total leakage through both cracks is expected to be less than 7 spm * (less than half that already present through the vent hole). [ The core spray system includes a design margin leakage ,

allowance the vent of approximately 100 gpm to allow for laakste through i

vessel nozzle. and thermal sleeve betvaan the toe-box and holes

No air bubble leakage was observed through the thermal aleeva during leakage is expected durine theaair bubble test, although some small TDCA Tha*a#--- *

• 4

_ , , - , - - - - - - - - ~ ~ - - _ _ _ _----- ~~ 1.h. Iltructural Integrity '

                                          'f           The core spray piping cracks at Peach Botton Unic 3 are expected to be caused by the influence of wald sensitisation or
                                              -        prior sensitisation of the core spray line material and subsequent enld work forming and installation. sources of                                          '

stress for possible Intergranular Stress Corrosion Cracking (ICSCC) are dependent on residual stresses from welding bending and deflection during installation. All identified stresses during normal reactor operation were found to be negligible. The loading considered include impingement load, seismic loading, pressure, weight and thermally induced loads. It is concluded that these normal operating loads do not result in stresses which are sufficient to cause the cracks observed in the piping. In the evaluation of crack arrest, the stresses due to bracket restraint and the fabrication residual stresses were also evaluated. Because the applied normal loading is predominantly displacement controlled, the stresses relax as the cracks grow. Crack180 about arrest is therefore expected when the crack grows to degreen. Stresses for stresses incurred during core spray injection are the design the piping. Design loadings include those listed above plus those resulting from system activation. It is " concluded that the structural integrity of the piping vill be maintained during core spray injection. In summary, the potential sources of stresses in the piping rasulting from fabrication, installation, normal operation, and caeration during postulated 1.oss of Coolant Accidents were reviewed. Potential causes of cracking and the likelihood of crack propogation were also evaluated. It is concluded that theconditions all structural of integrity of the piping will be maintained for operation. l.c. Emergency Core Cooling It has been concluded based on the analyses discussed below that there is no change in the Maximum Average planar Heat Cencration the presence Rate of (MAPLHCR) the cracks. limit of the upcoming Cycle 7 due to The estimated total leakage through the cracks is expected to be less than 7 gpm, compared to the rated flow, assumed in the ' reload 6250 spm. licensing analyses, through one cora spray system of (conservatively neglected in the licensing analyses) theCons effecta are negligible. -

                                       ,                                                However, to conservatively bound the Icakage CE has perforved a sensitivity study assuming that 10%

(or 625 gpm) of the core spray flow from the system with the cracks This is almost is lost from injecting into the reactor core region. 100 times the estimated leakage rate. I I

F, ., . . i' . e' i With this flow loss assumed, the bounding effect of the crack

 )                   -        on F.CCS Design        performance Basis           was evaluated by reanalyzing the limiting Accident (DBA)

I I LOCA event. The limiting break sito or single failure combination 'oss not changa with this nasumption. The original limiting LOCA event analysis for Reload 6, Cycle 7, resulted in a maximum PCT of 1954*F. The '

-                            analyses with an assumed 10% flow loss and all other assumptions identical to the reload analysia, results in a PCT of approximately 1960'F (a negligible change). This results in 240'F margin to the 10CFR50.46 limit of 2200*F. Since reanalysis indicates that sufficient margin remains, no reduction is required in the Cycle 7 MAPLHCR limits.

To further bound the effects of the cracks, the above analysis was repeated but with no credit for core spray heat transfer from the core spray system with the cracks. The resulting PCT increased to 2074*F, again well within the 2200'F PCT limit and again indicating that no reduction to MAPLHGR is required. 1.d. Structural Repair The above ECCS and structural analyses indicated that 1) the cracks are not expected to grow significantly beyond 180', 2) there is sufficient remaining structural integrity to accommodate the normal and injection loads, and 3) the cracks do not result in exceeding ECCS PCT limits. Nevertheless, to assure that the piping can accommodate any of the postulated loads two brackets will be velded across the piping arms and tee-box (see Figure 1). These brackets assure that the structural integrity of the piping is restored. The design basis ofennsiderations, ALARA the brackst wolds balanced strength and installation l.e Loose Parts Analysis Although it in anticipated that the piping will not break, particularly with the bracket design being installed, an evaluation piece of the possible consequences of a potential loose was performed. safety concerns: The evaluation addressed the following

1) Potential for corrosion or other chemical reaction to reactor materials; 2) Potential for flow blockage to a fuel bundle and subsequent for interference with control rod operation. fuel damage and; 3) Potential The probability action is zero. for unacceptable corrosion,or other chemical for the reactor vessel environment.The piping and bracket material are selected*

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