ML071640328

E-MAIL: (NPA-PD) Entergy LRA Letter Part 1 of 2

ML071640328
Person / Time
Site:	Pilgrim
Issue date:	05/01/2007
From:	Sanchez E Entergy Nuclear Operations
To:	Perry Buckberg Office of Nuclear Reactor Regulation
References
TAC MD3698
Download: ML071640328 (34)

Text

Page 1 of 1 Perry Buckberg - Entergy LRA letter part I of 2 From: "Sanchez, Edward" <esanchl@entergy.com>

To: <phbl@nrc.gov>

Date: 5/1/2007 7:22:38 PM

Subject:

Entergy LRA letter part 1 of 2

Perry, Attached is the first half of the 01 response letter. I split it in two due to file size.

Ed Sanchez Pilgrim Licensing file://C:\temp\GW}00002.HTM 5/31/2007

c:\te'm p\-G"W})0,0-00-2-.T-M-P Pg1 Mail Envelope Properties (4637CBA4.0E2 : 2 : 61666)

Subject:

Entergy LRA letter part 1 of 2 Creation Date 5/1/2007 7:21:18 PM From: "Sanchez, Edward" <esanchl (@entergy.com>

Created By: esanch I @&,entergy. cor Recipients nrc.gov OWGWPO01 .HQGWDO01 PHB I (Perry Buckberg)

Post Office Route OWGWPO01 .HQGWDO01 nrc.gov Files Size Date & Time MESSAGE 138 5/1/2007 7:21:18 PM TEXT.htm 1866 207027 (1 of 2).pdf 2862546 Mime.822 3921902 Options Expiration Date: None Priority: Standard ReplyRequested: No Return Notification: None Concealed

Subject:

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Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Stephen J. Bethay Director, Nuclear Assessment May 1,2007 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 License Renewal Application Amendment 16

REFERENCES:

1. Entergy Letter, License Renewal Application, dated January 25, 2006

2. NRC Safety Evaluation Report with Open Items Related to the Pilgrim License Renewal Application, dated March 2007

3. NRC Request for additional information for review of the Pilgrim license renewal application, dated March 26, 2007

4. Entergy Letter, Comments on NRC Draft Safety Evaluation Report Related to PNPS LRA, dated March 28, 2007 LETTER NUMBER: 2.07.027

Dear Sir or Madam:

In Reference 1, Entergy Nuclear Operations, Inc. applied for renewal of the Pilgrim Nuclear Power Station operating license. NRC TAC No. MC9669 was assigned to the application.

This letter provides information to address the Open Items from the NRC safety evaluation report (SER), (Reference 2). This letter also provides information in response to a request for additional information (Reference 3) related to Open Item 4.2. In addition, this letter includes LRA amendments resulting from review of the NRC SER.

Commitments made by this letter are contained in Attachment A.

Please contact Mr. Bryan Ford, (508) 830-8403, if you have questions regarding this subject.

I declare under penalty of perjury that the foregoing is true and correct. Executed on May 1,2007.

Sincerely, Stepen JM/ethay Director Nuclear Safety Assessment ERS/dI

Entergy Nuclear Operations, Inc. Letter Number: 2.07.027 Pilgrim Nuclear Power Station Page 2 Attachments:

Attachment A: Revised List of Regulatory Commitments Attachment B: Information in Response to the Open Items Listed in the Draft NRC SER, Including Associated LRA Amendments and License Condition Attachment C: LRA Amendments to Delete the BWRVIP-48 and BWRVIP-49 Fatigue Assessments as TLAAs Attachment D: Torus Room Concrete Base Mat Evaluation (Dr. Franz Ulm, M.I.T.)

Attachment E: Structural Integrity Associates Fluence Evaluation for PNPS cc: see next page

Entergy Nuclear Operations, Inc. Letter Number: 2.07.027 Pilgrim Nuclear Power Station Page 3 cc: with Attachments Mr. Perry Buckberg Mr. Joseph Rogers Project Manager Commonwealth of Massachusetts Office of Nuclear Reactor Regulation Assistant Attorney General U.S. Nuclear Regulatory Commission Division Chief, Utilities Division Washington, DC 20555-0001 1 Ashburton Place Boston, MA 02108 Alicia Williamson Mr. Matthew Brock, Esq.

Project Manager Commonwealth of Massachusetts Office of Nuclear Reactor Regulation Assistant Attorney General U.S. Nuclear Regulatory Commission Environmental Protection Division Washington, DQ. 20555-0001 One Ashburton Place Boston, MA 02108 Susan L. Uttal, Esq. Diane Curran, Esq.

Office of the General Counsel Harmon, Curran, and Eisenberg, L.L.P.

U.S. Nuclear Regulatory Commission 1726 M Street N.W., Suite 600 Mail Stop 0-15 D21 Washington, DC 20036 Washington, DC 20555-0001 Sheila Slocum Hollis, Esq. Molly H. Bartlett, Esq.

Duane Morris LLP 52 Crooked Lane 1667 K Street N.W., Suite 700 Duxbury, MA 02332 Washington, DC 20006 cc: without Attachments Mr. James S. Kim, Project Manager Mr. Robert Walker, Director Division of Operating Reactor Licensing Massachusetts Department of Public Health Office of Nuclear Reactor Regulation Radiation Control Program U. S. Nuclear Regulatory Commission Schrafft Center, Suite 1 M2A One White Flint North 4D9A 529 Main Street 11555 Rockville Pike Charlestown, MA 02129 Rockville, MD 20852 Mr. Jack Strosnider, Director Mr. Ken McBride, Director Office of Nuclear Material and Safeguards Massachusetts Energy Management Agency U.S. Nuclear Regulatory Commission 400 Worcester Road Washington, DC 20555-00001 Framingham, MA 01702 Mr. Samuel J. Collins, Administrator Mr. James E. Dyer, Director Region I . 4 Office of Nuclear Reactor Regulation

,U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission 475 Allendale Road Washington, DC 20555-00001 King of Prussia, PA 19406 NRC Resident Inspector Pilgrim Nuclear Power Station

ATTACHMENT A to Letter 2.07.027 (8 pages)

Revised List of Regulatory Commitments

Revised List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document.

Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section NoJ Comments Implement the Buried Piping and Tanks Inspection June 8, 2012 Letters B.1.2 / Audit Program as described in LRA Section B.1.2. 2.06.003 Item 320 and 2.06.057 2 Enhance the implementing procedure for ASME June 8, 2012 Letters B.1.6 / Audit Section XI inservice inspection and testing to specify 2.06.003 Item 320 that the guidelines in Generic Letter 88-01 or and approved BWRVIP-75 shall be considered in 2.06.057 determining sample expansion if indications are found in Generic Letter 88-01 welds.

3 Inspect fifteen (15) percent of the top guide locations As stated in the Letters B.1.8 / Audit using enhanced visual inspection technique, EVT-1, commitment. 2.06.003 Items 155, within the first 18 years of the period of extended and 320 operation, with at least one-third of the inspections to 2.06.057 be completed within the first six (6) years and at least and two-thirds within the first 12 years of the period of 2.06.064 extended operations. Locations selected for and examination will be areas that have exceeded the 2.06.081 neutron fluence threshold.

4 Enhance the Diesel Fuel Monitoring Program to June 8, 2012 Letters B.1.10/

include quarterly sampling of the security diesel 2.06.003 Audit Items generator fuel storage tank. Particulates (filterable and 320, 566 solids), water and sediment checks will be performed 2.06.057 on the samples. Filterable solids acceptance criteria and will be = 10 mg/l. Water and sediment acceptance 2.06.089 criteria will be = 0.05%.

5 Enhance the Diesel Fuel Monitoring Program to June 8, 2012 Letters B.1.10/

install instrumentation to monitor for leakage between 2.06.003 Audit Items the two walls of the security diesel generator fuel and 155, 320 storagbe tank to ensure that significant degradation is 2.06.057 not occurring.

6 Enhance the Diesel Fuel Monitoring Program to June 8, 2012 Letters B.1.10 /

specify acceptance criterion for UT measurements of 2.06.003 Audit Items emergency diesel generator fuel storage tanks and 165, 320 (T-126A&B). 12.06.057 I

1. COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section NoJ Comments 7 Enhance Fire Protection Program procedures to state June 8, 2012 Letters B.1.13.1 /

that the diesel engine sub-systems (including the fuel 2.06.003 Audit Items supply line) shall be observed while the pump is and 320, 378 running. Acceptance criteria will be enhanced to 2.06.057 verify that the diesel engine did not exhibit signs of and degradation while it was running; such as fuel oil, 2.06.064 lube oil, coolant, or exhaust gas leakage. Also, enhance procedures to clarify that the diesel-driven fire pump engine is inspected for evidence of corrosion in the intake air, turbocharger, and jacket water.system components as well as lube oil cooler.

The jacket water heat exchanger is inspected for evidence of corrosion or buildup to manage loss of material and fouling on the tubes. Also, the engine exhaust piping and silencer are inspected for evidence of internal corrosion or cracking.

8 Enhance the Fire Protection Program procedure for June 8, 2012 Letters B.1.13.1 /

Halon system functional testing to state that the 2.06.003 Audit Item Halon 1301 flex hoses shall be replaced if leakage and 320 occurs during the system functional test. 2.06.057 9 Enhance Fire Water System Program procedures to June 8, 2012 Letters B.1.13.2 /

include inspection of hose reels for corrosion. 2.06.003 Audit Item Acceptance criteria will be enhanced to verify no and 320 significant corrosion. 2.06.057 10 Enhance the Fire Water System Program to state that June 8, 2012 Letters B.1.13.2 /

a sample of sprinkler heads will be inspected using 2.06.003 Audit Item guidance of NFPA 25 (2002 Edition) Section and 320 5.3.1.1.1. NFPA 25 also contains guidance to repeat 2.06.057 this sampling every 10 years after initial field service testing.

11 Enhance the Fire Water System Program to state that June 8, 2012 Letters B.1.13.2 /

wall thickness evaluations of fire protection piping will 2.06.003 Audit Item be performed on system components using non- and 320 intrusive techniques (e.g., volumetric testing) to 2.06.057 identify evidence of loss of material due to corrosion.

These inspections will be performed before the end of the current operating term and at intervals thereafter during the period of extended operation. Results of the initial evaluations will be used to determine the appropriate inspection interval to ensure aging effects are identified prior to loss of intended function.

12 Implement the Heat Exchanger Monitoring Program June 8, 2012 Letters B.1.15 /

as described in LRA Section B.1.15. 2.06.003 Audit Item and 320 1_ _ 12.06..057 2

COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section NoJ Comments 13 Enhance the Instrument Air Quality Program to June 8, 2012 Letters B.1.17 /

include a sample point in the standby gas treatment 2.06.003 Audit Item and torus vacuum breaker instrument air subsystem and 320 in addition to the instrument air header sample points. .2.06.057 14 Implement the Metal-Enclosed Bus Inspection June 8, 2012 Letters B.1.18/

Program as described in LRA Section B.1.18. 2.06.003 Audit Item and 320 2.06.057 15 Implement the Non-EQ Inaccessible Medium-Voltage June 8, 2012 Letters B.1.19 /

Cable Program as described in LRA Section B.1.19. 2.06.003 Audit items Include developing a formal procedure to inspect and 311,320 manholes for in-scope medium voltage cable. 2.06.057 16 Implement the Non-EQ Instrumentation Circuits Test June 8, 2012 Letters 8.1.20 /

Review Program as described in LRA Section B.1.20. 2.06.003 Audit Item and 320 2.06.057 17 Implement the Non-EQ Insulated Cables and June 8, 2012 Letters B.1.21 /

Connections Program as described in LRA Section 2.06.003 Audit Item B.1.21. and 320 2.06.057 18 Enhance the Oil Analysis Program to periodically June 8, 2012 Letters B.1.22/

change CRD pump lubricating oil. A particle count 2.06.003 Audit Item and check for water will be performed on the drained and 320 oil to detect evidence of abnormal wear rates, 2.06.057 contamination by moisture, or excessive corrosion.

19 Enhance Oil Analysis Program procedures for June 8, 2012 Letters B.1.22 /

security diesel and reactor water cleanup pump oil 2.06.003 Audit Item changes to obtain oil samples from the drained oil. and 320 Procedures for lubricating oil analysis will be 2.06.057 enhanced to specify that a particle count and check for water are performed on oil samples from the fire water jump diesel, security diesel, and reactor water cleanup pumps.

20 Implement the One-Time Inspection Program as June 8, 2012 Letters B.1.23 /

described in LRA Section B.1.23. 2.06.003 Audit Items and 219,320 2.06.057 and 2.07.023 21 Enhanrce the Periodic Surveillance and Preventive June 8, 2012 Letters B.1.24 /

Maintenance Program as necessary to assure that 2.06.003 Audit Item the effects of aging will be managed as described in and 320 LRA Section B.1.24. _ 2.06.057

3

COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section No./

Comments 22 Enhance the Reactor Vessel Surveillance Program to June 8, 2012 Letters B.1.26 /

proceduralize the data analysis, acceptance criteria, 2.06.003 Audit Item and corrective actions described in LRA Section and 320 B. 1.26. 2.06.057, 23 Implement the Selective Leaching Program in June 8, 2012 Letters B.1.27/

accordance with the program as described in LRA 2.06.003 Audit Item Section B.1.27. and 320 2.06.057 24 Enhance the Service Water Integrity Program June 8, 2012 Letters B.1.28 /

procedure to clarify that heat transfer test results are 2.06.003 Audit Item trended, and 320 2.06.057 25 Enhance the Structures Monitoring Program June 8, 2012 Letters B.1.29.2 /

procedure to clarify that the discharge structure, 2.06.003 Audit Items security diesel generator building, trenches, valve and 238, 320 pits, manholes, duct banks, underground fuel oil tank 2.06.057 foundations, manway seals and gaskets, hatch seals and gaskets, underwater concrete in the intake structure, and crane rails and girders are included in the program. In addition, the Structures Monitoring Program will be revised to require opportunistic inspections of inaccessible concrete areas when they become accessible.

26 Enhance Structures Monitoring Program guidance for June 8, 2012 Letters B.1.29.2 /

performing structural examinations of elastomers 2.06.003 Audit Item (seals, gaskets, seismic joint filler, and roof and 320 elastomers) to identify cracking and change in 2.06.057 I material properties.

27 Enhance the Water Control Structures Monitoring June 8, 2012 Letters B.1.29.3 /

Program scope to include the east breakwater, jetties, 2.06.003 Audit Item and onshore revetments in addition to the main and 320 breakwater. 2,06.057 28 Enhance System Walkdown Program guidance June 8, 2012 Letters B.1.30/

documents to perform periodic system engineer 2.06.003 Audit Items inspections of systems in scope and subject to aging and 320, 327 management review for license renewal in 2.06.057 accordance with 10 CFR 54.4(a)(1) and (a)(3).

Inspections shall include areas surrounding the subject systems to identify hazards to those systems.

Inspections of nearby systems that could impact the subject systems will include SSCs that are in scope and subject to aging management review for license renewal in accordance with 10 CFR 54.4(a)(2).

4

COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section No./

Comments 29 Implement the Thermal Aging and Neutron Irradiation June 8, 2012 Letters B.1.31 /

Embrittlement of Cast Austenitic Stainless Steel 2.06.003 Audit Items (CASS) Program as described in LRA Section B.1.31. and 257, 320 2.06.057 30 Perform a code repair of the CRD return line nozzle June 30, 2015 Letter B.1.3 / Audit to cap weld if the installed weld repair is not approved 2.06.057 Items 141, via accepted code cases, revised codes, or an 320 approved relief request for subsequent inspection intervals. I I I 31 At lea~st 2 years prior to entering the period of June 8, 2012 Letters 4.3.3 / Audit extended operation, for the locations identified in 2.06.057 Items 302, NUREG/CR-6260 for BWRs of the PNPS vintage, and 346 PNPS will implement one or more of the following: June 8, 2010 for 2.06.064 (1) Refine the fatigue analyses to determine valid CUFs submitting the and less than 1 when accounting for the effects of reactor water aging 2.06.081 environment. This includes applying the appropriate Fen management and factors to valid CUFs determined in accordance with one of program if PNPS 2.07.005 the following: selects the 1.,-For locations, including NUREG/CR-6260 locations, with option of existing fatigue analysis valid for the period of extended managing the operation, use the existing CUF to determine the affects of aging environmentally adjusted CUF. due to

2. More limiting PNPS-speciiic locations with a valid CUF environmentally may be added in addition to the NUREG/CR-6260 locations. assisted fatigue.

3. Representative CUF values from other plants, adjusted to or enveloping the PNPS plant specific external loads may be used ifdemonstrated applicable to PNPS.

4.* An analysis using an NRC-approved version of the ASME code of NRC-approved alternative (e.g., NRC-approved code case) may be performed to determine a valid CUF.

The determination of Fen will account for operating times with both hydrogen water chemistry and normal water chemistry.

(2) Manage the effects of aging due to fatigue at the affected locations by an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC).

(3) Repair or replace the affected locations before exceeding a CUF of 1.0.

Should PNPS select the option to manage the aging effects due to environmental-assisted fatigue during the period of extended operation, details of the aging management program such as scope, qualification, method, and frequency will be submitted to the NRC at least 2.years prior to the period of extended operation.

5

COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section No./

Comments 32 Implement the enhanced Bolting Integrity Program June 8, 2012 Letters Audit items described in Attachment C of Pilgrim License 2.06.057 364, 373, Renewal Application Amendment 5 (Letter 2.06.064). and 389, 390, 2.06.064 432,443, and 470 2.06.081 33 PNPS will inspect the inaccessible jet pump thermal As stated in the Letter Audit Items sleeve and core spray thermal sleeve welds if and commitment. 2.06.057 320, 488 when the necessary technique and equipment become available and the technique is demonstrated by the vendor, including delivery system.

34 Within the first 6 years of the period of extended June 8, 2018 Letters Audit Items operation and every 12 years thereafter, PNPS will 2.06.057 320, 461 inspect the access hole covers with UT methods. and Alternatively, PNPS will inspect the access hole 2.06.089 covers in accordance with BWRVIP guidelines should such guidance become available.

35 At least 2 years prior to entering the period of June 8, 2012 Letters Audit Item extended operation, for reactor vessel components, June 8, 2010 for 2.06.057 345 including the feedwater nozzles, PNPS will implement submitting the and one o( more of the following: aging 2.06.064 (1) Refine the fatigue analyses to determine valid management and CUFs less than 1. Determine valid CUFs based on program if PNPS 2.06.081 numbers of transient cycles projected to be valid selects the for the period of extended operation. Determine option of CUFs in accordance with an NRC-approved managing the version of the ASME code or NRC-approved affects of aging.

alternative (e.g., NRC-approved code case).

(2) Manage the effects of aging due to fatigue at the affected locations by an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC).

(3) Repair of replace the affected locations before exceeding a CUF of 1.0.

Should PNPS select the option to manage the aging effects due to fatigue during the period of extended operation, details of the aging management program such as scope, qualification, method, and frequency will be submitted to the NRC at least 2 years prior to the period of extended operation.

6

COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section NoJ Comments 36 To ensure that significant degradation on the bottom June 8, 2012 Letter Audit Items of the condensate storage tank is not occurring, a 2.06.057 320, 363 one-time ultrasonic thickness examination in accessible areas of the bottom of the condensate storage tank will be performed. Standard examination and sampling techniques will be utilized.

37 The BWR Vessel Internals Program includes June 8, 2012 Letter A.2.1.8/

inspections of the steam dryer. Inspections of the 2.06.089 Conference steam dryer will follow the guidelines of BWRVIP-139 call on and General Electric SIL 644 Rev. 1. September 1 25, 2006 38 Enhance the Diesel Fuel Monitoring Program to June 8, 2012 Letter B.1.10/

include periodic ultrasonic thickness measurement of 2.06.089 Audit Item the bottom surface of the diesel fire pump day tank. 565 The first ultrasonic inspection of the bottom surface of the diesel fire pump day tank will occur prior to the period of extended operation, following engineering analysis to determine acceptance criteria and test locations. Subsequent test intervals will be determined based on the first inspection results.

39 Perfo[m a one-time inspection of the Main Stack June 8, 2012 Letter B.1.23 /

foundation prior to the period of extended operation. 2.06.094 Audit Item 581 40 Enhance the Oil Analysis Program by documenting June 8, 2012 Letter B.1.22 /

program elements 1 through 7 in controlled 2.06.094 Audit Items documents. The program elements will include 553 and 589 enhancements identified in the PNPS license renewal application and subsequent amendments to the application. The program will include periodic sampling for the parameters specified under the Parameters Monitored/Inspected attribute of NUREG-1801 Section XI.M39, Lubricating Oil Analysis. The controlled documents will specify appropriate acceptance criteria and corrective actions in the event acceptance criteria are not met. The basis for acceptance criteria will be defined.

41 Enhance the Containment Inservice Inspection (CII) June 8, 2012 Letter A.2.1.17 and Progrqm to require augmented inspection in 2.06.094 B.1.16.1 accordance with ASME Section XI IWE-1240, of the drywell shell adjacent to the sand cushion following I indications of water leakage into the annulus air gap.

42 Implement the Bolted Cable Connections Program, June 8, 2012 Letter A.2.1.40 and described in Attachment C of Pilgrim License 2.07.003 B.1.34 Renewal Application 11 (Letter 2.07.003), prior to the period of extended operation.

7

1. COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section No./

Comments 43 Include within the Structures Monitoring Program June 8, 2012 Letter A.2.1.32 and provisions to ensure groundwater samples are 2.07.005 B.1.29.2 evaluated periodically to assess the aggressiveness of groundwater to concrete, as described in Attachment E of LRA Amendment 12 (Letter 2.07.005), prior to the period of extended operation.

44 Perform another set of the UT measurements just As stated in the Letter A.2.1.17 and above and adjacent to the sand cushion region prior commitment. 2.07.010 B.1.16.1 to the period of extended operation and once within the first 10 years of the period of extended operation.

45 If groundwater continues to collect on the Torus As stated in the Letters A.2.1.32 and Room floor, obtain samples and test such water to commitment. 2.07.010 B.1.29.2 determine its pH and verify the water is non- and aggressive as defined in NUREG-1 801 Section IllI.A1 2.07.027 item III.A.1-4 once prior to the period of extended operation and once within the first ten years of the period of extended operation.

46 Inspect the condition of a sample of the torus hold- June 8, 2012 Letter A.2.1.32 and down bolts and associated grout and determine 2.07.027 B.1.29.2 appropriate actions based on. the findings prior to the period of extended operation.

47 Submit to the NRC an action plan to improve Sept.15, 2007 Letter 4.2.2, benchmarking data to support approval of new P-T 2.07.027 A.2.2.1.1, curves for Pilgrim. and A.2.2.1.2 48 On or before June 8, 2010, Entergy will submit to the June 8, 2010 Letter 4.2, 4.7.1, NRC calculations consistent with Regulatory 2.07.027 A.1.1 and Guide 1.190 that will demonstrate limiting fluence A.2.2.1 values will not be reached during the period of extended operation.

8 4

ATTACHMENT B to Letter 2.07.027 (17 pages)

Information in Response to the Open Items Listed in the Draft NRC SER, Including Associated LRA Amendments and License Condition

01 2.3.3.6: (SER Section 2.3.3.6 - Security Diesel)

LRA Table 2.3.3-6 shows the component types subject to an AMR but the security diesel system was not in the FSAR or in any license renewal drawings; therefore, the staff could not determine the portion of the security diesel system within the scope of license renewal.

Additionally, the staff could not determine whether any components within the scope of license renewal were not shown as subject to an AMR. The staff referred this issue to NRC Region I who will determine whether security diesel system components are within the scope of license renewal.

01 2.3.3.6 Response Entergy provided NRC Region I with support as requested.

207027 Page 1 of 17

01 3.0.3.2.10: (SER Section 3.0.3.2.10 - Fire Protection Program)

The applicant is taking an exception to the GALL Report program element "detection of aging effects," specifically:

The NUREG-1 801 program states that approximately 10 percent of each type of penetration seal should be visually inspected at least once every refueling outage. The PNPS program specifies inspection of approximately 20 percent of the seals each operating cycle, with all accessible fire barrier penetration seals being inspected at least once every five operating cycles.

The LRA states that, because aging effects typically are manifested over several years, this variation in inspection frequency is insignificant. GALL AMP XI.M26 specifies approximately 10 percent of each type of seal should be inspected visually at least every refueling outage (two years). The applicant clarified that the program specifies inspection of approximately 20 percent of the seals, including at least one seal of each type, each operating cycle, with all accessible fire barrier penetration seals being inspected at least once every five operating cycles. The applicant needs to address how to manage the aging effect of inaccessible fire barrier penetration seals.

O 3.0.3.2.10 Response The PNPS requirement to inspect penetration seals applies to 100% of the seals. The word "accessible" is not necessary in the discussion of the exception for Detection of Aging Effects in the PNPS program. All fire barrier penetration seals are inspected at least once every five operating cycles. In LRA Appendix B, Section B.1.13.1, the word "accessible" is removed resulting in the following description of the exception for Detection of Aging Effects.

The NUREG-1 801 program states that approximately 10% of each type of penetration seal should be visually inspected at least once every refueling outage. The PNPS program specifies inspection of approximately 20% of the seals each operating cycle, with all aGceeesble fire barrier penetration seals being inspected at least once every five operating cycles.

207027 Page 2 of 17

01 3.0.3.3.2: (SER Section 3.0.3.3.2 - Containment Inservice Insoection and Section 3.5.2.2.1 -

PWR and BWR Containments)

A recent NRC Region 1 inspection team observations indicated the following:

• The flow switch in the bellows rupture drain had failed its surveillance in December 2005 and has not been fixed or evaluated. In addition, the flow switch also failed in 1999.

• Monitoring of other drains has been inconclusive and not well documented.

" The torus room floor has had water on the floor on multiple occasions.

In Request for'Additional Information (RAI) B.1.16.1, dated November 7, 2006, the applicant was asked to address the above finding and discuss the impact on the aging management of potential loss of material due to corrosion in the inaccessible area of the Mark I steel containment drywell shell, basemat, including the sand pocket region for the period of extended operation. -

01 3.0.3.3.2 Response Entergy letter dated March 13, 2007 provided information to address this open item and RAI B.1.16.1. With~regard to the issue of water on the torus room floor, Attachment D to this letter contains a report prepared by a consultant to Entergy that provides a detailed evaluation of the groundwater seepage through the concrete basemat.

Commitments 43, 45, and 46 will be implemented to address this issue. Commitment 45 made in the March 13, 2007 Entergy letter is revised by this letter to require it be performed once within the first ten years of the period of extended operation in addition to it being performed once prior to the period of extended operation. Commitment 46 is added by this letter. These commitments are listed in Attachment A to this letter and read as follows:

43 Include within the Structures Monitoring Program provisions to ensure groundwater samples are evaluated periodically to assess the aggressiveness of groundwater to concrete, as described in Attachment E of LRA Amendment 12 (Letter 2.07.005), prior to the period of extended operation.

45 If groundwater continues to collect on the Torus Room floor, obtain samples and test such water to determine its pH and verify the water is non-aggressive as defined in NUREG-1 801 Section IllI.A1 item III.A.1-4 once prior to the period of extended operation and once within the first ten years of the period of extended operation.

46 Inspect the condition of a sample of the torus hold-down bolts and associated grout and determine appropriate actions based on the findings prior to the period of extended operation.

207027 Page 3 of 17

OI 4.2: (SER Sections: 3.0.3.2.15 - Reactor Vessel Surveillance Proqram, 4.2 - Reactor Vessel Neutron Embrittlement, 4.7.1 - Reflood Thermal Shock of the Reactor Vessel Internals. 4.7.2.1 BWRVIP-05, Reactor Vessel Circumferential Welds)

Due to the lack of benchmarking data in support of the plant-specific RAMA fluence calculations, the staff finds neutron fluence values unacceptable for use in the reactor vessel neutron embrittlement TLAAs.

01 4.2 Response 01 4.2 was clarified by the NRC in a request for additional information (RAI) transmitted in a letter dated March 26, 2007. The RAIs and responses are provided below.

RAI# 4.2

1. Fluence' was calculated for the Pilgrim reactor vessel (RV) for the extended 60-year licensed operating period (54 effective full power years (EFPY).of facility operation),

using the Radiation Analysis Modeling Application (RAMA) fluence methodology. The RAMA fluence methodology was previously approved by the NRC staff, and the results are acceptable for licensing actions provided that: (1) the RAMA application follows the guidance in Regulatory Guide 1.190 and (2) RV fluence calculations have at least one credible plant-specific surveillance capsule for benchmarking.

The applicant provided 54 EFPY fluence values for the Pilgrim RV beltline materials in Section 4.2.1 of the License Renewal Application (LRA). These fluence values were used throughout Section 4.2 of the LRA for the RV neutron embrittlement time limited aging analyses (TLAAs). However, due to the lack of a credible plant-specific benchmark, the staff finds the 54 EFPY fluence values provided in LRA Section 4.2.1 unacceptable for use in the RV neutron embrittlement TLAAs. Therefore, the staff requests that the applicant revise Section 4.2.1 of the LRA to provide an acceptable neutron fluence evaluation or an alternative proposal for closing this TLAA topic in the LRA review.

2. Due to the lack of benchmarking data in support of the plant-specific RAMA fluence calculations, the staff cannot complete its review of the TLAAs in LRA Sections 4.2.2, 4.2.3, 4.2.4, 4.2.5, 4.2.6 and 4.7.1, as well as the aging management program (AMP) on the RV material surveillance program, using the current fluence values for the Pilgrim RV that were provided in LRA Section 4.2.1. Therefore, the staff requests that the applicant revise LRA Sections 4.2.2, 4.2.3,4.2.4, 4.2.5, 4.2.6, 4.7.1, and the AMP on the RV material surveillance program to provide an acceptable evaluation of these topics or an alternative proposal for closing these topics in the LRA review.

Response

The benchmarking validation of the RAMA fluence calculation is ongoing for the Pilgrim reactor vessel and internals. The RAMA calculated fluence is approximately 56% of the benchmark fluence calculated from the available surveillance capsule dosimetry. Uncertainties between the calculated and measured results from the dosimetry are still being examined to determine a possible cause for the discrepancy. To ensure resolution of this issue, Commitment 47, which reads as follows, is added by this letter.

207027 Page 4 of 17

47 On or before September 15, 2007 submit to the NRC an action plan to improve benchmarking data to support approval of new P.T curves for Pilgrim.

To address this issue, an alternative analysis is provided as a means to close this TLAA topic in the LRA review. To address fluence-related TLAAs for the period of extended operation, Entergy has evaluated the affected TLAAs to determine the limiting fluence value. The evaluation included information presented in LRA sections 4.2.1, 4.2.2, 4.2.3, 4.2.4, 4.2.5, 4.2.6, 4.7.1, and the AMP on the RV material surveillance program. From this evaluation the limiting fluence was determined.

The alternative analysis to determine the limiting fluence value is included as Attachment E.

This analysis assumes increasing fluence levels until an ASME Code or regulatory limit is reached based on the projected changes in material properties. Changes in the vessel (ferritic) steel material properties are measured by an increase in adjusted reference temperature or a decrease in Chiarpy upper shelf energy. The effects of increasing fluence on the austenitic stainless steel core shroud and internals was also considered. By assuming increasing fluence levels, the analysis identifies the maximum fluence that can be experienced while meeting the Code and regulatory criteria. This analysis also shows that there is a large margin available to this limiting fluence at the end of the period of extended operation.

The analysis determined that the limiting fluence value was set by the temperature required to perform the ASME Code hydrostatic test. The temperature to perform the hydrostatic test is prescribed by ASME Section Xl, Article G-2400 that requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed increasing fluence and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum starting temperature for the hydrostatic test was set at 212°F with a corresponding maximum allowable 1/4T fluence of 4.12E+18 n/cm 2 .

If fluence remains below this limiting value during the period of extended operation, the fluence will result in acceptable results for all fluence-related TLAAs. To confirm that the limiting fluence will not be reached during the period of extended operation and consequently that all of the fluence-relatedTLAAs remain valid, Commitment 48, which reads as follows, is added by this letter.

On or before June 8, 2010, Entergy will submit to the NRC calculations consistent 48 with Regulatory Guide 1.190 that will demonstrate limiting fluence values will not be reached during the period of extended operation.

Entergy would find it acceptable ifthis commitment became a license condition.

It should be noted that at the ACRS meeting on April 4, 2007, reference was made to EPRI research that investigated the irradiated behavior of stainless steel components in order to predict service life. Further review has shown that the predictions of service life related to fluence are not directly relevant in this case. The core shroud and the top guide are components that are susceptible to aging effects. However, a review of the analyses related to the core shroud found that the only time-limited aging analysis (TLAA) involves the fatigue analysis and calculation of cumulative usage factors (CUFs) for the shroud repair. The core shroud does not 207027 Page 5 of 17

affect the operating P-T limit curves and there is no criterion on fluence that would further limit the operation of the core shroud structure. Similarly, the top guide does not affect the operating P-T limit curves, and there is no criterion on fluence that would further limit the operation of the top guide structure.

PNPS has re-evaluated the neutron embrittlement issues of Sections 4.2 and 4.7.1 and prepared revised LRA sections below. The Reactor Vessel Material Surveillance Program, with the changes to the fluence extrapolation, is correct as written, and no changes to Appendix B, Section B.1.26 are necessary.

LRA Amendments 4.2 REACTOR VESSEL NEUTRON EMBRI'TLEMENT The regulations governing reactor vessel integrity are in 10 CFR 50. Section 50.60 requires that all light-water reactors meet the fracture toughness, pressure-temperature limits, and material surveillance program requirements for the reactor coolant pressure boundary as set forth in 10 CFR 50 Appendiices G and H.

The PNPS current licensing basis analyses evaluating reduction of fracture toughness of the PNPS reactor vessel for 40 years are TLAA. The reactor vessel neutron embrittlement TLAA has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii) as summarized below. Fifty-four effective full-power years (EFPY) are projected for the end of the period of extended operation (60 years) assuming an average capacity factor of 90% for 60 years.

4.2.1 Reactor Vessel Fluence Calculated fluence is based on a time-limited assumption defined by the operating term. As such, fluence is the time-limited assumption for the time-limited aging analyses that evaluate reactor vessel neutron embrittlement.

Fluence values were calculated using the RAMA fluence methodology. The RAMA fluence methodology was developed for the Electric Power Research Institute, Inc. and the boiling water reactor vessel and internals project (BWRVIP) for the purpose of calculating neutron fluence in boiling water reactor components. This methodology has been approved by the NRC (Reference 4.2-20) for application in accordance with Regulatory Guide (RG) 1.190; assuming:

the results are appropriately benchmarked.

The benchmarking validation of the RAMA fluence calculation is ongoing for the Pilgrim reactor vessel. The RAMA calculated fluence is approximately 56% of the benchmark fluence calculated from the available surveillance capsule dosimetry. Uncertainties between the calculated and measured results from the dosimetry are still being examined to determine a possible cause for the discrepancy. Commitment 47 requires a plan for resolving this discrepancy to be developed and submitted for review by September 2007.

An alternative analysis to determine the limiting fluence value has been performed. This analysis assumes increasing fluence levels until an ASME Code or regulatory limit is reached based on the projected changes in material properties. Changes in the vessel (ferritic) steel material properties are measured by an increase in adjusted reference temperature or a decrease in Charpy upper shelf energy. The effects of increasing fluence on the austenitic stainless steel core shroud and internals was also considered. By assuming increasing fluence 207027

.Page 6 of 17

levels, the analysis identifies the maximum fluence that can be experienced while meeting the Code and regulatory criteria.

The analysis determined that the limiting fluence value was set by the temperature required to perform the ASME Code hydrostatic test. The temperature to perform the hydrostatic test is prescribed by ASME Section Xl, Article G-2400, which requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed different fluences and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum temperature starting for the hydrostatic test was set at 212°F The corresponding maximum allowable fluence is a 1/4T fluence of 4.12E+18 n/cm 2 . This fluence level was the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will result in acceptable results for all fluence-related TLAAs. Commitment 48 is to confirm that the limiting fluence will not be reached during the period of extended operation and consequently that all of the fluence-related TLAAs will be valid to the end of the period of extended operation.

At PNPS, the limiting beltline material for 40 years consists of 6 plates and their connecting welds, all adjacent to the active fuel zone. No nozzles are included in the limiting beltline materials for the current term of operation (Reference 4.2-2).

The beltline will be re-evaluated for 60 years. An evaluation of the RTNDT for nozzle forgings and welds is expected to show that their adjusted reference temperature at 54 EFPY will be well below the adjusted reference temperatures used in determining the P-T limits. Thus, the nozzle forgings and welds are not expected to be the limiting items for the period of extended operation.

4.2.2 Pressure-Temperature Limits Appendix G of 10 CFR 50 requires that reactor vessel boltup, hydrotest, pressure tests, normal operation, and anticipated operational occurrences be accomplished within established pressure-temperature (P-T) limits. These limits are established by calculations that utilize the materials and fluence data obtained through the Reactor Vessel Surveillance Program.

Pilgrim received License Amendment 227 dated March 29, 2007 that extended the existing P-T limit curves for Pilgrim through Cycle 18.

The P-T limit ciirves will continue to be updated, as required by Appendix G of 10 CFR Part 50 or as operational needs dictate. This updating will assure that the operational limits remain valid through the period of extended operation. Maintaining the P-T limit curves in accordance with Appendix G of 10 CFR 50 assures that the effects of aging on the intended function(s) will be adequately managed for the period of extended operation consistent with 10 CFR 54.21 (c)(1)(iii).

4.2.3 Charpy Upper-Shelf Energy Appendix G of 10 CFR 50 requires that reactor vessel beltline materials "have Charpy upper-shelf energy ... of no less than 75 ft-lb initially and must maintain Charpy upper-shelf energy throughout the life of the vessel of no less than 50 ft-lb ..... The initial (unirradiated) values of 207027 Page 7 of 17

upper-shelf energy (CVUSE) for PNPS beltline welds were provided to the NRC in correspondence responding to Generic Letter 92-01 (References 4.2-9, 4.2-10).

Regulatory Guide 1.99, Radiation Embrittlementof Reactor Vessel Materials,Revision 2, provides two methods for determining Charpy upper-shelf energy (CVUSE). Position 1 applies for material that does not have surveillance data and Position 2 applies for material with surveillance data. Position 2 requires a minimum of two sets of credible material surveillance data. Since PNPS has data from only one material surveillance capsule, Position 2 does not apply. For Position 1, the percent drop in CVUSE for a stated copper content and neutron fluence is determined by reference to Figure 2 of Regulatory Guide 1.99, Revision 2. This percentage drop is applied to the initial CVUSE to obtain the adjusted CVUSE.

The predictions for percent drop in CVUSE at 54 EFPY must be based on chemistry data, the maximum 1/4T fluence values, and unirradiated CVUSE data submitted to the NRC in the PNPS response to GL 92-01. The predicted CVUSE values for 54 EFPY will utilize Regulatory Guide 1.99 Position 1. The predictions will use Regulatory Guide 1.99, Position 1, Figure 2; specifically, the formula for the lines will be used to calculate the percent drop in CVUSE (Reference 4.2-14).

PNPS will use chemistry data from previous licensing submittals, the PNPS response to GL 92-01 (References 4.2-9, 4.2-10, 4.2-14), and the 1/4T fluence values to be determined to perform linear interpolation on the CVUSE percent drop values in RG 1.99, Revision 2, Figure 2.

The license renewal SER for BWRVIP-74 (Reference 4.2-11), Action Item #10, states that each license renewal applicant shall demonstrate that the percent reduction in Charpy USE for their beltline materials is less than that specified for the limiting BWR/3-6 plates and the non-Linde 80 submerged arc welds given in BWRVIP-74. This action item is not applicable to PNPS ifthe PNPS projected CVUSE remains above the 50 ft-lb limit, even for the period of extended operation.

An analysis determined that the limiting fluence value is the fluence that corresponds to the maximum temperature limit for performing the ASME Code hydrostatic test. The temperature to perform the hydrostatic test is prescribed by ASME Section XI, Article G-2400, which requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed different fluences and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum temperature for the hydrostatic test was set at 212 0 F. The corresponding maximum allowable fluence is a 1/4T fluence of 4.12E18 n/cm 2 . This fluence is the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will result in acceptable results for the reactor vessel Charpy upper shelf energy TLAA. To confirm that the limiting fluence will not be reached during the period of extended operation and consequently that this TLAA will be valid to the end of the period of extended operation, Commitment 48 is added.

4.2.4 Adjusted Reference Temperature Irradiation by high-energy neutrons raises the value of RTNDT for the reactor vessel. RTNDT is the reference temperature for nil-ductility transition as defined in Section NB-2320 of the ASME Code. The initial RTNDT is determined through testing of unirradiated material specimens. The shift in reference temperature, ARTNDT, is the difference in the 30 ft-lb index temperatures from the average Charpy curves measured before and after irradiation. The adjusted reference 207027 Page 8 of 17

temperature (ART) is defined as initial RTNDT + ARTNDT + margin. The margin is defined in RG 1 .99, Revision'2. The P-T curves are developed from the ART value for the vessel materials.

RG 1.99 Revision 2 defines the calculation methods-for RTNOT and ART.

The PNPS reactor vessel-was evaluated for an assumed exposure of less than 1019 nvt of neutrons with energies exceeding 1 MeV (Reference 4.2-1). After approximately 4.17 EFPY, the first surveillance capsule was withdrawn from the vessel and tested. The capsule test report concludes that the shift in RTNDT and upper-shelf energy over 32 EFPY will be within 10 CFR 50 guidelines.

PNPS will project values for ARTNDT and ART at 54 EFPY using the methodology of RG 1.99.

These values will be calculated using the chemistry data, margin values, initial RTNDT values, and chemistry factors (CFs) contained in the PNPS response to GL 92-01 (References 4.2-3, 4.2-9, 4.2-10, 4.2-13). Initial RTNDT values are from report SIR-00-082, which was submitted in 2001 as part of the PNPS P-T limit change request (Reference 4.2-5). The 1/4T fluence values discussed in Section 4.2.1 will be used. New fluence factors (FFs) will be calculated using the expression in RG 1.99, Revision 2, Equation 2, where the fluence factor is given by FF=f (0.28-0.lO*Iogf)

In this equation, f is the 1/4T fluence value. The new ARTNDT values will be calculated by multiplying the CF and the FF for each plate and weld. Calculated margins and the initial RTNDT will then be added to the calculated ARTNDT in order to arrive at the new value of ART.

An analysis determined that the limiting fluence value is the fluence that corresponds to the maximum temperature limit for performing the ASME Code hydrostatic test. The temperature to perform the hydrostatic test is prescribed by ASME Section XI, Article G-2400, which requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed different fluences and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum temperature for the hydrostatic test was set at 212 0 F. The corresponding maximum allowable fluence is a 1/4T fluence of 4.12E18 n/cm 2. This fluence is the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will result in acceptable results for the reactor vessel adjusted reference temperature TLAA. To confirm that the limiting fluence will not be reached during the period of extended operation and consequently that this TLAA will be valid to the end of the period of extended operation, Commitment 48 is added.

4.2.5 Reactor Vessel Circumferential Weld Inspection Relief Relief from reactor vessel circumferential weld examination requirements under Generic Letter 98-05 is based on an analysis indicating acceptable probability of failure per reactor operating year. The analysis is based on reactor vessel metallurgical conditions as well as flaw indication sizes and frequencies of occurrence that are expected at the end of a licensed operating period.

PNPS received NRC approval for this relief for the remainder of the original 40-year license term. The basis for this relief request is an analysis that satisfied the limiting conditional failure probability for the circumferential welds at the expiration of the current license, based on BWRVIP-05 and the extent of neutron embrittlement (References 4.2-16, 4.2-17). The 207027 Page 9 of 17

anticipated changes in metallurgical conditions expected over the extended operating period require additional analysis to extend this relief request.

The NRC evaluation of BWRVIP-05 utilized the FAVOR code to perform a probabilistic fracture mechanics (PFM) analysis to estimate the reactor pressure vessel (RPV) shell weld failure' probabilities. Three key inputs to the PFM analysis are (1) the estimated end-of-life mean neutron fluence, (2) mean chemistry values based on vessel types, and (3) the assumption of potential for beyond-design-basis events.

PNPS will compare the reactor vessel limiting circumferential weld parameters to those used in the NRC analyis for the first two key assumptions. The data will be from the NRC SER for PNPS Relief Request 28 (Reference 4.2-17), and from the data in Table 2.6.4 of the NRC SER for BWRVIP-05 (Reference 4.2-18). (For comparison, the EOL mean RTNDT will be calculated without margin and hence will be lower than the Section 4.2.2 RTNDT value.)

The procedures and training used to limit cold over-pressure events will be the same as those approved by the NRC when PNPS requested' approval of the BWRVIP-05 technical alternative for the current license term.

An analysis determined that the limiting fluence value is the fluence that corresponds to the maximum temlerature limit for performing the ASME Code hydrostatic test. The temperature to perform the hydrostatic test is prescribed by ASME Section XI, Article G-2400, which requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed different fluences and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum temperature for the hydrostatic test was set at 212 0 F. The corresponding maximum allowable fluence is a 1/4T fluence of 4.12E18 n/cm 2. This fluence is the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will result in acceptable results for the reactor vessel circumferential weld failure probability TLAA. To confirm that the limiting fluence will not be reached during the period of extended operation and consequently that this TLAA will be valid to the end of the period of extended operation, Commitment 48 is added.

4.2.6 Reactor Vessel Axial Weld Failure Probability The BWRVIP recommendations for inspection of reactor vessel shell welds (BWRVIP-05) are based on generic analyses supporting an NRC SER conclusion that the generic-plant axial weld failure rate is no more than 5x10.8 per reactor year (Reference 4.2-18). BWRVIP.05 showed that this axial weld failure rate is orders of magnitude greater than the 40-year end-of-life circumferential weld failure probability, and used this analysis to justify relief from inspection of the circumferential welds as described above.

PNPS received relief from the circumferential weld inspections for the remainder of the original 40-year operating term (Reference 4.2-17). The basis for this relief request was a plant-specific analysis that showed the limiting conditional failure probability for the PNPS circumferential welds at the end of the original operating term was less than the values calculated in the BWRVIP-05 SER (Reference 4.2-11). The BWRVIP-05 SER concluded that the reactor vessel failure frequency due to failure of the limiting axial welds in the BWR fleet at the end of 40 years of operation is less than 5x1 06 per reactor year. This failure frequency is dependent upon given assumptions of flaw density, distribution, and location. The failure frequency also assumes that 207027 Page 10 of 17

"essentially 100%" of the reactor vessel axial welds will be inspected. The PNPS relief request requires additional relief request if less than 90% coverage is achieved.

PNPS will compare the reactor vessel limiting axial weld parameters to those used in the NRC analysis. The parameters used will be those from the NRC SER for BWRVIP-05 (Reference 4.2-18) from the NRC Supplemental SER for BWRVIP-05 (Reference 4.2-19).

The supplemental SER required the limiting axial weld to be compared with data found in Table 3 of the document. Originally, the supplemental SER identified PNPS as a limiting plant for the BWR fleet; however, in the discussion it is noted that the high EOL value of RTNDT for PNPS calculated by the BWRVIP is due to the use of an initial RTNDT of 0°F. The supplemental SER notes that the docketed value of initial RTNOT (from the RVID) is -48 0 F, and therefore the EOL value of RTNDT for PNPS is not bounding for the BWR fleet. The supplemental SER stated that the axial welds for the Clinton plant are the limiting welds for the BWR fleet and vessel failure probability determined for Clinton should bound the BWR fleet.

The limiting values will be compared to the values assumed in the analysis performed by the NRC staff in the BWRVIP-05 supplemental SER and the 64 EFPY limits and values obtained from Table 2.6- 5 of the SER. As such, this TLAA will be projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii).

An analysis determined that the limiting fluence value is the fluence that corresponds to the maximum temperature limit for performing the ASME Code hydrostatic test. The temperature to perform the hydrostatic test is prescribed by ASME Section XI, Article G-2400, which requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed different fluences and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum temperature for the hydrostatic test was set at 212 0 F. The corresponding maximum allowable fluence is a 1/4T fluence of 4.12E1 8 n/cm 2. This fluence is the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will result in acceptable results for the reactor vessel axial weld failure probability TLAA. To confirm that the limiting fluence will not be reached during the period of extended operation and consequently that this TLAA will be valid to the end of the period of extended operation, Commitment 48 is added.

4.7.1 Reflood Thermal Shock of the Reactor Vessel Internals UFSAR Section 3.3.6.8 addresses reflood thermal shock of the reactor vessel internals (core shroud). This evaluation of thermal shock was considered a TLAA as it is potentially based on shroud material properties that are affected by neutron fluence.

The shroud material is Type 304 stainless steel, which is not significantly affected by irradiation.

An analysis determined that the limiting fluence value is the fluence that corresponds to the maximum temperature limit for performing the ASME Code hydrostatic test. The temperature to perform the hydrostatic test is prescribed by ASME Section Xl, Article G-2400, which requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed different fluences and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum temperature for the 207027 Page 11 of 17

hydrostatic test was set at 212 0 F. The corresponding maximum allowable fluence is a 1/4T fluence of 4.12E18 n/cm 2. This fluence level was the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will result in acceptable results for the reflood thermal shock TLAA. To confirm that the limiting fluence will not be reached during the period of extended operation and consequently that this TLAA will be valid to the end of the period of extended operation, Commitment 48 is added.

Changes to existinq UFSAR Section 3.3.6.8 information oresented in Section A.1.1 of the LRA (oaqe A-3) are revised as follows:

3. Shroud inner surfaces at highest irradiation zone The mo.sti point on the innor surface Of the shroud is subj.ctcd t- a total integratod neutron flX of 2.7 x 1020 RA. (ý,1 MeV) by the end of station lif The peak thermal shock stress is 155,700 psi, corresponding to a peak strain of 0.57 percent. The shroud material is Type 304 stainless steel, which is not significantly affected by irradiation. The material does experience a loss in reduction of area. Because reduction of area is the property which determines tolerable local strain, irradiation effects can be neglected. The peak strain resulting from thermal shock at the inside of the shroud represents no loss of integrity of the reactor vessel inner volume. The creVio limit of Typo 304 stainless steol is approached at a fiuonco of 8 x 1021 Nm2 (BWRVII 36). As the PNPS shroud Will remai, bolw .. that ..u. ne, V. l for tho per*id of extended operation, the shroud Will remain Sorvicoable.

UFSAR Supplement Sections are revised to read as follows:

A.2.2.1.1 Reactor Vessel Fluence Calculated fluence is based on a time-limited assumption defined by the operating term. As such, fluence is the time-limited assumption for the time-limited aging analyses that evaluate reactor vessel embrittlement. Fluence values were calculated using the RAMA fluence calculation method. The RAMA fluence method was developed for the Electric Power Research Institute, Inc. and the Boiling Water Reactor Vessel and Internals Project (BWRVIP) for the purpose of calculating neutron fluence in boiling water reactor components. This method has been approvedcby the NRC (Reference A.2-9) for application in accordance with Regulatory Guide 1.190 provided the fluence calculations for the reactor are appropriately benchmarked.

The benchmarking validation of the RAMA fluence calculation is ongoing for the PNPS reactor vessel. The RAMA calculated fluence is approximately 56% of the benchmark fluence calculated from the available surveillance capsule dosimetry. Uncertainties between the calculated and measured results from the dosimetry are still being examined to determine a possible cause for the discrepancy. An action plan to improve benchmarking data to support approval of new P-T curves will be developed and submitted for NRC review.

An alternative analysis to determine the limiting fluence value has been performed (Reference A.2-12). This analysis assumes increasing fluence levels until an ASME Code or regulatory limit is reached based on the projected changes in material properties. Changes in the vessel (ferritic) steel material properties are measured by an increase in adjusted reference temperature or a decrease in Charpy upper shelf energy. The effects of increasing fluence on the austenitic stainless steel core shroud and internals was also considered. By assuming increasing fluence levels, the analysis identifies the maximum fluence that can be experienced while meeting the Code and regulatory criteria.

207027 Page 12 of 17

The analysis determined that the limiting fluence value was set by the temperature required to perform the ASME Code hydrostatic test. The temperature to perform the hydrostatic test is prescribed by ASME Section Xl, Article G-2400, which requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed different fluences and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum temperature starting for the hydrostatic test was set at 212°F The corresponding maximum allowable fluence is a 1/4T fluence of 4.12E+18 n/cm 2. This fluence level was the limiting fluence value identified.

On or before June 8, 2010, Entergy will submit to the NRC calculations consistent with Regulatory Guide 1.190 that will demonstrate limiting fluence values will not be reached during the period of eXtended operation.

A.2.2.1.2 Pressure-Temperature Limits Appendix G of 10 CFR 50 requires that reactor vessel boltup, hydrotest, pressure tests, normal operation, and anticipated operational occurrences be accomplished within established pressure-temperature (P-T) limits. These limits are established by calculations that utilize the materials and fluence data obtained through the Reactor Vessel Surveillance Program.

Pilgrim received License Amendment 227 dated March 29, 2007 that extended the existing P-T limit curves for Pilgrim through Cycle 18.

The P-T limit curves will continue to be updated, as required by Appendix G of 10 CFR Part 50 or as operational needs dictate. This updating will assure that the operational limits remain valid through the period of extended operation. Maintaining the P-T limit curves in accordance with Appendix G of 10 CFR 50 assures that the effects of aging on the intended function(s) will be adequately managed for the period of extended operation consistent with 10 CFR 54.21 (c)(1)(iii).

A.2.2.1.3 Charpy Upper-Shelf Energy Appendix G of 10 CFR 50 requires that reactor vessel beltline materials "have Charpy upper-shelf energy ... of no less than 75 ft-lb initially and must maintain Charpy upper-shelf energy throughout the life of the vessel of no less than 50 ft-lb...." The initial (unirradiated) values of upper-shelf energy (CvUSE) for PNPS beltline welds were provided to the NRC in correspondence responding to Generic Letter 92-01.

Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, provides two methods for determining Charpy upper-shelf energy (CvUSE). Position 1 applies for material that does not have surveillance data and Position 2 applies for material with surveillance data. Position 2 requires a minimum of two sets of credible material surveillance data. Since PNPS has data from only one material surveillance capsule, Position 2 does not apply. For Position 1, the percent drop in CvUSE for a stated copper content and neutron fluence is determined by reference to Figure 2 of Regulatory Guide 1.99, Revision 2. This percentage drop is applied to the initial CvUSE to obtain the adjusted CvUSE.

The predictions for percent drop in CVUSE at 54 EFPY must be based on chemistry data, the maximum 1/4T fluence values, and unirradiated CvUSE data submitted to the NRC in the PNPS response to GL 92-01. The predicted CvUSE values for 54 EFPY will utilize Regulatory Guide 1.99 Position 1. The predictions will use Regulatory Guide 1.99, Position 1, Figure 2; specifically, the formula for the lines will be used to calculate the percent drop in CvUSE.

207027 Page 13 of 17

PNPS will use chemistry data from previous licensing submittals, the PNPS response to GL 92-01, and the 1/4T fluence values to be determined to perform linear interpolation on the CvUSE percent drop values in RG 1.99, Revision 2, Figure 2.

The license renewal SER for BWRVIP-74, Action Item #10, states that each license renewal applicant shall demonstrate that the percent reduction in Charpy USE for their beltline materials is less than that specified for the limiting BWR/3-6 plates and the non-Linde 80 submerged arc welds given in BWRVIP-74. This action item is notapplicable to PINPS if the PNPS projected CvUSE remains above the 50 ft-lb limit, even for the period of extended operation.

An analysis determined that the limiting fluence value is the fluence that corresponds to the maximum temperature limit for performing the ASME Code hydrostatic test. The temperature to perform the hydrostatic test is prescribed by ASME Section XI, Article G-2400, which requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed different fluences and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum temperature for the hydrostatic test was set at 212 0 F. The corresponding maximum allowable fluence is a 1/4T fluence of 4.12E18 n/cm 2. This fluence is the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will result in acceptable results for the reactor vessel Charpy upper shelf energy TLAA. To confirm that this TLAA will be valid to the end of the period of extended operation, Entergy will submit to the NRC on or before June 8, 2010 calculations consistent with Regulatory Guide 1.190 that will demonstrate limiting fluence values will not be reached during the period of extended operation.

A.2.2.1.4 Adjusted Reference Temperature Irradiation by high-energy neutrons raises the value of RTNOT for the reactor vessel. RTNDT is the reference temperature for nil-ductility transition as defined in Section NB-2320 of the ASME Code. The initial RTNDT is determined through testing of unirradiated material specimens. The shift in reference temperature, ARTNDT, is the difference in the 30 ft-lb index temperatures from the average Charpy curves measured before and after irradiation. The adjusted reference temperature (ART) is defined as initial RTNDT + ARTNDT + margin. The margin is defined in RG 1.99, Revision 2. The P-T curves are developed from the ART value for the vessel materials.

RG 1.99 Revision 2 defines the calculation methods for RTNDT and ART.

The PNPS reactor vessel was evaluated for an assumed exposure of less than 1019 nvt of neutrons with energies exceeding 1 MeV. After approximately 4.17 EFPY, the first surveillance capsule was withdrawn from the vessel and tested. The capsule test report concludes that the shift in RTNDT and upper-shelf energy over 32 EFPY will be within 10 CFR 50 guidelines.

PNPS will project values for ARTNDT and ART at 54 EFPY using the methodology of RG 1.99.

These values will be calculated using the chemistry data, margin values, initial RTNDT values, and chemistry factors (CFs) contained in the PNPS response to GL 92-01. Initial RTNDT values are from report SIR-00-082, which was submitted in 2001 as part of the PNPS P-T limit change request. The 1/4T fluence values discussed in Section 4.2.1 will be used. New fluence factors (FFs) will be calculated using the expression in RG 1.99, Revision 2, Equation 2, where the fluence factor is given by FF = f (0.28-0. 10*logf) 207027 Page 14 of 17

In this equation, f is the 1/4T fluence value. The new ARTNDT values will be calculated by multiplying the CF and the FF for each plate and weld. Calculated margins and the initial RTNDT will then be added to the calculated ARTNDT in order to arrive at the new value of ART.

An analysis determined that the limiting fluence value is the fluence that corresponds to the maximum temperature limit for performing the ASME Code hydrostatic test. The temperature to perform the hydrostatic test is prescribed by ASME Section Xl, Article G-2400, which requires a safety factor of-.1.5 on the pressure stress interisity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed different fluences and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum temperature for the hydrostatic test was set at 212 0 F. The corresponding maximum allowable fluence is a 1/4T fluence of 4.12E18 n/cm 2 . This fluence is the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will result in acceptable results for the reactor vessel adjusted reference temperature TLAA. To confirm that this TLAA will be valid to the end of the period of extended operation, Entergy will submit to the NRC on or before June 8, 2010 calculations consistent with Regulatory Guide 1.190 that will demonstrate limiting fluence values will not be reached during the period of extended operation.

A.2.2.1.5 Reactor Vessel Circumferential Weld Inspection Relief Relief from reactor vessel circumferential weld examination requirements under Generic Letter 98-05 is based on an analysis indicating acceptable probability of failure per reactor operating year. The analysis is based on reactor vessel metallurgical conditions as well as flaw indication sizes and frequencies of occurrence that are expected at the end of a licensed operating period.

PNPS received NRC approval for this relief for the remainder of the original 40-year license term. The basis for this relief request is an analysis that satisfied the limiting conditional failure probability for the circumferential welds at the expiration of the current license, based on BWRVIlP-05 and the extent of neutron embrittlement. The anticipated changes in metallurgical conditions expected over the extended operating period require additional analysis to extend this relief request.

The NRC evaluation of BWRVIP-05 utilized the FAVOR code to perform a probabilistic fracture mechanics (PFM) analysis to estimate the reactor pressure vessel (RPV) shell weld failure probabilities. Three key inputs to the PFM analysis are (1) the estimated end-of-life mean neutron fluence, (2) mean chemistry values based on vessel types, and (3) the assumption of potential for beyond-design-basis events.

PNPS will compare the reactor vessel limiting circumferential weld parameters to those used in the NRC analysis for the first two key assumptions. The data will be from the NRC SER for PNPS Relief Request 28, and from the data in Table 2.6.4 of the NRC SER for BWRVIP-05.

(For comparison, the EOL mean RTNbT will be calculated without margin and hence will be lower than the Section 4.2.2 RTNDT value.)

The procedures and: training used to limit cold over-pressure events will be the same as those approved by the NRC when PNPS requested approval of the BWRVIP-05 technical alternative for the current license term.

An analysis determined that the limiting fluence value is the fluence that corresponds to the maximum temperature limit for performing the ASME Code hydrostatic test. The temperature to 207027 Page 15 of 17

perform the hydrostatic test is prescribed by ASME Section Xl, Article G-2400, which requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed different fluences and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum temperature for the hydrostatic test was set at 2120F. The corresponding maximum allowable fluence is a 1/4T fluence of 4.12E18 n/cm 2. This fluence is the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will result in acceptable results for the reactor vessel circumferential weld failure probability TLAA. To confirm that this TLAA will be valid to the end of the period of extended operation, Entergy will submit to the NRC on or before June 8, 2010 calculations consistent with Regulatory Guide 1.190 that will demonstrate limiting fluence values will not be reached during the period of extended operation.

A.2.2.1.6 Reactor Vessel Axial Weld Failure Probability The BWRVIP recommendations for inspection of reactor vessel shell welds (BWRVIP-05) are based on generic analyses supporting an NRC SER conclusion that the generic-plant axial weld failure rate is no more than 5x106 per reactor year. BWRVIP-05 showed that this axial weld failure rate is orders of magnitude greater than the 40-year end-of-life circumferential weld failure probability, and used this analysis to justify relief from inspection of the circumferential welds as described above.

PNPS received relief from the circumferential weld inspections for the remainder of the original 40-year operating term. The basis for this relief request was a plant-specific analysis that showed the limiting conditional failure probability for the PNPS circumferential Welds at the end of the original operating term was less than the values calculated in the BWRVIP-05 SER. The BWRVIP-05 SER concluded that the reactor vessel failure frequency due to failure of the limiting axial welds in the BWR fleet at the end of 40 years of operation is less than 5x10B per reactor year. This failure frequency is dependent upon given assumptions of flaw density, distribution, and location. The failure frequency also assumes that "essentially 100%" of the reactor vessel axial welds will be inspected. The PNPS relief request requires additional relief request if less than 90% coverage is achieved.

PNPS will compare the reactor vessel limiting axial weld parameters to those used in the NRC analysis. The parameters used will be those from the NRC SER for BWRVIP-05 from the NRC Supplemental SER for BWRVIP-05.

The supplemental SER required the limiting axial weld to be compared with data found in Table 3 of the document. Originally, the supplemental SER identified PNPS as a limiting plant for the BWR fleet; however, in the discussion it is noted that the high EOL value of RTNDT for PNPS calculated by the BWRVIP is due to the use of an initial RTNDT of 00 F. The supplemental SER notes that the docketed value of initial RTNDT (from the RVID) is -48 0 F, and therefore the EOL value of RTNDT for PNPS is not bounding for the BWR fleet. The supplemental SER stated that the axial welds for the Clinton plant are the limiting welds for the BWR fleet and vessel failure probability determined for Clinton should bound the BWR fleet.

The limiting values will be compared to the values assumed in the analysis performed by the NRC staff in the BWRVIP-05 supplemental SER and the 64 EFPY limits and values obtained from Table 2.6- 5 of the SER. As such, this TLAA will be projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii).

207027 Page 16 of 17

An analysis determined that the limiting fluence value is the fluence that corresponds to the maximum temperature limit for performing the ASME Code hydrostatic test. The temperature to perform the hydrostatic test is prescribed by ASME Section XI, Article G-2400, which requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracture of the vessel during this test. The vessel integrity analysis assumed different fluences and calculated the corresponding hydrostatic test temperatures. As fluence increases, higher temperatures are required to perform the test to meet the ASME Code criteria. The maximum temperature for the hydrostatic test was set at 212 0 F. The corresponding maximum allowable fluence is a 1/4T fluence of 4.12E18 n/cm 2. This fluence is the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will result in acceptable results for the reactor vessel axial weld failure probability TLAA. To confirm that this TLAA will be valid to the end of the period of extended operation, Entergy will submit to the NRC on or before June 8, 2010 calculations consistent with Regulatory Guide 1.190 that will demonstrate limiting fluence values will not be reached during the period of extended operation.

The following referernce is added to UFSAR Supplement Section A.2.3.

A.2-12 Bethay, Stephen J. (Entergy), to Document Control Desk (NRC), "License Renewal Application Amendment 16," letter 2.07.027 dated May 1, 2007, Attachment E, Structural Integrity Associates Fluence Evaluation for PNPS.

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ATTACHMENT C to Letter 2.07.027 (1 page)

LRA Amendments to Delete the BWRVIP-48 and BWRVIP-49 Fatigue Assessments as TLAAs

LRA Amendments to Delete BWRVIP-48 and BWRVIP-49 Fatigue Assessments as TLAAs Sections 4.7.2.2 and 4.7.2.3 of the draft SER discuss BWRVIP-48, Vessel ID Attachment Welds and BWRVIP-49, Instrument Penetrations, respectively. Both sections appropriately conclude that the BWRVIP analysis does not constitute a TLAA for license renewal since the CLB does not include a plant-specific 40-year CUF calculation for these components.

Section 4.7.2.3 states, "In a letter dated May 11, 2006, the applicant amended the LRA to delete the BWRVIP-49 fatigue assessment for the RPV IPNs as a TLAA for the LRA. The amendment deleted LRA Section A.2.2.6." The referenced amendment letter does not contain the stated change. Apparently the change was made to the Vermont Yankee LRA, but due to administrative 'Oversight was not made to the PNPS LRA. The change is equally applicable to PNPS. Amendments to the LRA to delete the BWRVIP-49 fatigue assessment as a TLAA are provided.

Also, Section 4.7.2.2 should have a similar discussion of changes to the LRA necessary to delete discussion of BWRVIP-48 as a TLAA. Amendments to the LRA to delete the BWRVIP-48 fatigue assessment as a TLAA are provided.

LRA Amendments to Delete the BWRVIP-48 Fatigue Assessment as a TLAA Delete Section 4.7.2.2, "BWRVIP-48, Vessel ID Attachment Welds."

Delete entry for BWRVIP-48, vessel ID attachment welds fatigue analysis from Table 4.1-1, "List of PNPS TLAA and Resolution."

Delete entry for cracking - fatigue with TLAA-metal fatigue from ID attachment welds in Table 3.1.2-1. Cracking managed by the BWR Vessel ID Attachment Welds Program remains in the table.

Delete Section A.2.2.5, "Vessel ID Attachment Welds Fatigue Analysis."

LRA Amendments to Delete the 'BWRVIP-49 Fatigue Assessment as a TLAA Delete Section 4.7.2.3, "BWRVIP-49, Instrument Penetrations."

Delete entry for BWRVIP-49, instrument penetrations fatigue analysis from Table 4.1-1, "List of PNPS TLAA and Resolution."

Delete entry for cracking - fatigue with TLAA-metal fatigue from nozzles, reactor vessel instrumentation (N15, N16) in Table 3.1.2-1. Cracking managed by the BWR Penetrations Program remains in the table.

Delete Section A.2.2.6, "Instrument Penetrations Fatigue Analysis."