ML071640347

E-MAIL: (NPA-PD) Pilgrim Amendment 17 of 5/17/2007

ML071640347
Person / Time
Site:	Pilgrim
Issue date:	05/18/2007
From:	Perry Buckberg NRC/NRR/ADRO/DLR/RLRA
To:
References
TAC MD3698
Download: ML071640347 (55)

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Page 1 of 1 Perry Buckberg - Pilgrim Amendment 17 of 5/17/2007 From:

Perry Buckberg To:

Pilgrim Staff Date:

5/18/2007 7:02:29 AM

Subject:

Pilgrim Amendment 17 of 5/17/2007 Gentlemen, The attached letter was e-mailed last night. It's revised Open Item 4.2 response supercedes that of the last amendment, it revises Commitment 45 and it revises line items in tables 3.1.1 and 3.2.1 related to reduction of fracture toughness.

Thanks, Perry Buckberg Project Manager - Division of License Renewal phone: (301)415-1383 fax:

(301)415-3031 phbl@nrc.gov file://C:\\temp\\GW}00002.HTM 5/31/2007

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Subject:

Creation Date From:

Created By:

Pilgrim Amendment 17 of 5/17/2007 5/18/2007 7:02:29 AM Perry Buckberg PHB 1 @nrc.gov Recipients nrc.gov OWGWPO01.HQGWDOO1 AM RCA CC (Rajender Auluck) nrc.gov OWGWPOO2.HQGWDOO1 AM EBF (Edwin Forrest)

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MAM4 (Matthew Mitchell) nrc.gov TWGWPO02.HQGWDOOI Action Delivered Date & Time 5/18/2007 7:02:36 Delivered 5/18/2007 7:02:36 Delivered 5/18/2007 7:02:41 Delivered 5/18/2007 7:02:36 Delivered 5/18/2007 7:02:38 Delivered 5/18/2007 7:02:37

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C:\\t em p \\-G-W}10'-0 0-2--.TVM-P Page311 To Be Delivered:

Status Tracking:

Immediate Delivered

Is Enteigy Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Stephen J. Bethay Director, Nuclear Assessment May 17, 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 17

REFERENCES:

LETTER NUMBER:

Dear Sir or Madam:

1.

Entergy Letter, License Renewal Application, dated January 25, 2006 (TAC MC9669)

2.

NRC Request for additional information for review of the Pilgrim License Renewal Application, dated March 26, 2007

3.

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

4.

Entergy Letter, License Renewal Application Amendment 16, dated May 1, 2007 2.07.029 In Reference 1, Entergy Nuclear Operations, Inc. applied for renewal of the Pilgrim Nuclear Power Station operating license.

Attachment A provides a revised listing of regulatory commitments. Attachments B and C provide additional information associated with the application. Attachments D and E provide a revised response to the request for additional information (RAI) in Reference 2 associated with Open Item 4.2 in the draft NRC safety evaluation report related to the Pilgrim license renewal application (LRA) (Reference 3). The revised RAI response supersedes the response to Open Item 4.2 in Reference 4.

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 17, 2007.

Sincerely, Stephen J Bethay Director Nuclear Safety Assessment ERS/dI

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Letter Number: 2.07.029 Page 2 Attachments:

A:

Revised List of Regulatory Commitments B:

Revision to Commitment 45 C:

LRA Amendments to Revise LRA Tables 3.1.1-55 and 3.2.1-3 D:

Revised Response to the NRC Request for Additional Information Related to Open Item 4.2 E:

Structural Integrity Associates Fluence Evaluation for PNPS cc: with Attachments Mr. Perry Buckberg Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Alicia Williamson Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Susan L. Uttal, Esq.

Office of the General Counsel U.S. Nuclear Regulatory Commission Mail Stop 0-15 D21 Washington, DC 20555-0001 Sheila Slocum Hollis, Esq.

Duane Morris LLP 1667 K Street N.W., Suite 700 Washington, DC 20006 cc: without Attachments Mr. James S. Kim, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission One White Flint North O-8C2 11555 Rockvilleý Pike Rockville, MD 20852 Mr. Jack Strosnider, Director Office of Nuclear Material and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-00001 Mr. Samuel J. Collins, Administrator Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 NRC Resident Inspector Pilgrim Nuclear Power Station Mr. Joseph Rogers Commonwealth of Massachusetts Assistant Attorney General Division Chief, Utilities Division 1 Ashburton Place Boston, MA 02108 Mr. Matthew Brock, Esq.

Commonwealth of Massachusetts Assistant Attorney General Environmental Protection Division One Ashburton Place Boston, MA 02108 Diane Curran, Esq.

Harmon, Curran, and Eisenberg, L.L.P.

1726 M Street N.W., Suite 600 Washington, DC 20036 Molly H. Bartlett, Esq.

52 Crooked Lane Duxbury, MA 02332 Mr. Robert Walker, Director Massachusetts Department of Public Health Radiation Control Program Schraft Center, Suite 1 M2A 529 Main Street Charlestown, MA 02129 Mr. Ken McBride, Director Massachusetts Energy Management Agency 400 Worcester Road Framingham, MA 01702 Mr. James E. Dyer, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-00001 4

ATTACHMENT A to Letter 2.07.029 (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 Xl 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/I. 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 storage 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).

2.06.0571 Page 1 of 8

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 testingr.

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 2.06.057 Page 2 of 8

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 B.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 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 ari performed on oil samples from the fire water pump 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 Enhance 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_

Page 3 of 8

COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section NoJ 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 1

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 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.

2.06057 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).

Page 4 of 8 4

COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section NoJ 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

31 At least 2 years prior to entering the period of extended operation, for the locations identified in NUREG/CR-6260 for BWRs of the PNPS vintage, PNPS will implement one or more of the following:

(1) Refine the fatigue analyses to determine valid CUFs less than 1 when accounting for the effects of reactor water environment. This includes applying the appropriate Fen factors to valid CUFs determined in accordance with one of the follbwing:

1. For locations, including NUREG/CR-6260 locations, with existing fatigue analysis valid for the period of extended operation, use the existing CUF to determine the environmentally adjusted CUF.

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

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

June 8, 2012 June 8, 2010 for submitting the aging management program if PNPS selects the option of managing the affects of aging due to environmentally assisted fatigue.

Letters 2.06.057 and 2.06.064 and 2.06.081 and 2.07.005 4.3.3 / Audit Items 302, 346 Page 5 of 8

COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section NoJ 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 or 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.

Page 6 of 8

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 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 Perform 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 Program 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 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.

Page 7 of 8

COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section NoJ 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 room As stated in the Letters A.2.1.32 and 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-1801 Section IllI.A1 2.07.027 item III.A.1-4 once prior to the period of extended and operation and once every five years during the period 2.07.029 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.

Page 8 of 8

ATTACHMENT B to Letter 2.07.029 (1 page)

Revision to Commitment 45

Revision to Commitment 45 Entergy letter dated March 13, 2007 added Commitment 45. Entergy letter dated May 1, 2007 revised Commitment 45 as part of the response to Open Item 3.0.3.3.2. Commitment 45 is revised to require performance once every five years during the period of extended operation in addition to once prior to the period of extended operation. This revised commitment is listed in Attachment A to this letter and reads as follows:

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-1801 Section II.A1 item III.A.1-4 once prior to the period of extended operation and once every five years during the period of extended operation.

Page 1 of 1

ATTACHMENT C to Letter 2.07.029 (1 page)

LRA Amendments to Revise LRA Tables 3.1.1-55 and 3.1.2-3

LRA Amendments to Revise LRA Tables 3.1.1-55 and 3.1.2-3 LRA Table 3.1.1-55 on Page 3.1-30 includes a line item for Cast austenitic stainless steel Class 1 pump casings, and valve bodies and bonnets exposed to reactor coolant >2500C (>4820F). For this line item, the discussion column is modified to read as follows.

The Inservice Inspection Program or the One-Time Inspection Program manage the reduction of fracture toughness in cast austenitic stainless steel components of the reactor coolant pressure boundary.

LRA Table 3.1.2'3 on Page 3.1-72 includes a line item for valve bodies < 4" NPS, with CASS material and with the aging effect reduction of fracture toughness. For this line item, the aging management program column is modified to add Inservice Inspection to One-Time Inspection. All other entries for this line item remain unchanged.

Page 1 of 1

ATTACHMENT D to Letter 2.07.029 (12 pages)

Revised Response to the NRC Request for Additional Information Related to Open Item 4.2

01 4.2: (SER Sections: 3.0.3.2.15 - Reactor Vessel Surveillance Program, 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 embrittjement 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 RAI and response is 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, was added by Entergy letter dated May 1, 2007.

Page 1 of 12

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 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. 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 maximum mean RTNDT value for the vessel axial welds of 1 140F to remain below a calculated reactor vessel failure frequency of 5x1 0B per reactor-year. The corresponding maximum allowable ID fluence for the axial welds was determined to be 3.37E+18 n/cm 2.

If fluence remains below this limiting value during the period of extended operation, the fluence will yield 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-related TLAAs remain valid, Commitment 48, which reads as follows, was added by Entergy letter dated May 1, 2007.

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 if this 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 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.

Page 2 of 12

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 Amendment 4.2 REACTOR VESSEL NEUTRON EMBRITrLEMENT 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 Appendices 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 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 is set by the maximum mean RTNOT value for the vessel axial welds of 1 140F to remain below a calculated reactor vessel failure Page 3 of 12

frequency of 5x10.8 per reactor-year. The corresponding maximum allowable ID fluence for the axial welds was determined to be 3.37E+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 yield 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 6nticipated 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).

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 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 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 PNRS 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.

Page 4 of 12

The predictions for percent drop in CvUSE at 54 EFPY must be based on chemistry data, the maximum 1/4t1fluence 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, and 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 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 set by the maximum mean RTNDT value for the vessel axial welds of 114°F to remain below a calculated reactor vessel failure frequency of 5x1 0- per reactor-year. The corresponding maximum allowable ID fluence for the axial welds was determined to be 3.37E+18 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 yield 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, 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 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 (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 Page 5 of 12

FF = f (0.28-0.1 0*logf)

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 set by the maximum mean RTNDT value for the vessel axial welds of 114 0F to remain below a calculated reactor vessel failure frequency of 5x108 per reactor-year. The corresponding maximum allowable ID fluence for the axial welds was determined to be 3.37E+18 n/cm2. This fluence is the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will yield 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 aswell 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 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) the 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 (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 bythe 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 set by the maximum mean RTNDT value for the vessel axial welds of 1 140F to remain below a calculated reactor vessel failure frequency of 5x10- per reactor-year. The corresponding maximum allowable ID fluence for the axial welds was determined to be 3.37E+1 8 n/cm 2. This fluence is the limiting fluence value identified.

Page 6 of 12

If fluence remains below this limiting value during the period of extended operation, the fluence will yield 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 5x10.6 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 4.2-18). 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 operatidn is less than 5x10-6 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.

Applicant Action Item 12 from the NRC SER for BWRVIP-74 specified that applicants should monitor axial beltline weld embrittlement. One acceptable method was to determine that the mean RTNDT of the limiting axial beltline weld at the end of the period of extended operation is less than the values specified in Table 1 of the FSER for BWRVIP-74. The limiting mean RTNDT value of 1 140F for the axial. welds was determined to be equivalent to a failure frequency of less than 5x10.6 per reactor-year.

An analysis determined that the ID fluence value that yields a mean RTNDT value for the vessel axial welds of 11 40F is 3.37E+1 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 Section43.3.6.8 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 set by the maximum mean RTNDT value for the vessel axial welds of 114 0F to remain below a calculated reactor vessel failure frequency of 5x1 06 per reactor-year. The corresponding maximum allowable ID fluence for the axial welds was determinedbto be 3.37E+1 8 n/cm2. This fluence level is the limiting f luence value identified.

Page 7 of 12

If fluence remains below this limiting value during the period of extended operation, the fluence will yield 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 presented in Section A.1.1 of the LRA (pacqe A-3) are revised as follows:

3. Shroud inner, surfaces at highest irradiation zone The FRoct irradiatod point on the innre curt6 9i o the Shroud is cubjoctod to a total intogratod n.utron f*lux f 2.7 x 1 R%" (> 1 Mo")

by the end of,tation

,if.. 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. T4he e......

limit of Typo 301 ctainl..

steel is app..a.hod at a Nflunc of 8 x 1021 n/cm2 (BWRVIP 36). As the PNPS chrlud Will rmain b-*

olw that Iluonco loYal foFr the pr;,d of evfend, operation, the

.hrovud Will romain,.o 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 approved by 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.

Page 8 of 12

The analysis determined that the limiting fluence value is set by the maximum mean RTNDT value for the vessel axial welds of 1 140F to remain below a calculated reactor vessel failure frequency of 5x10' per reactor-year. The corresponding maximum allowable ID fluence for the axial welds was determined to be 3.37E+1 8 n/cm 2. This fluence level is 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 bolt-up, hydrostatic tests, 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.

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.

Page 9 of 12

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 not applicable to PNPS 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 set by the maximum mean RTNDT value for the vessel axial welds of 1 140F to remain below a calculated reactor vessel failure frequency of 5x10,6 per reactor-year. The corresponding maximum allowable ID fluence for the axial welds was determined to be 3.37E+18 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 yield 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 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 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 1t4T 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, and Equation 2, where the fluence factor is given by FF = f (0.28-0.1 O*1ogf In this equation, f is the 1/4T fluence value. The new ARTNOT 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 set by the maximum mean RTNDT value for the vessel axial welds of 11 4°F to remain below a calculated reactor vessel failure frequency of 5x10B per reactor-year. The corresponding maximum allowable ID fluence for the axial welds was determined to be 3.37E+18 n/cm 2. This fluence is the limiting fluence value identified.

'Page 10 of 1.2

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 BWRVIP-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) the 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 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 set by the maximum mean RTN[DT value for the vessel axial welds of 11 40F to remain below a calculated reactor vessel failure frequency of 5x10,6 per reactor-year. The corresponding maximum allowable ID fluence for the axial welds was determined to be 3.37E+18 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 yield 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.

Page II of 12

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 5x10 6 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 5x1 0e 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.

Applicant Action Item 12 from the NRC SER for BWRVIP-74 specified that applicants should monitor axial beltline weld embrittlement. One acceptable method was to determine that the mean RTNDT of the limiting axial beltline weld at the end of the period of extended operation is less than the values specified in Table 1 of the FSER for BWRVIP-74. The limiting mean RTNDT value of 1 140F for the axial welds was determined to be equivalent to a failure frequency of less than 5x1 0,6 per reactor-year.

An analysis determined that the ID fluence value that yields a mean RTNDT value for the vessel axial welds of 1 14°F is 3.37E+18 n/cm2. This fluence is the limiting fluence value identified.

If fluence remains below this limiting value during the period of extended operation, the fluence will yield 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 reference 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 17," letter 2.07.029 dated May 17, 2007, Attachment E, Structural Integrity Associates Fluence Evaluation for PNPS.

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ATTACHMENT E to Letter 2.07.029 (22 pages)

Structural Integrity Associates Fluence Evaluation for PNPS

V Structural Integrity Associates, Inc.

File No.: PNPS-27Q-301 CALCULATION PACKAGE Project No.: PNPS-27Q PROJECT NAME: Evaluation of Fluence Issues for Pilgrim Nuclear Power Station CONTRACT NO.: 4500555337, Rev. 1 CLIENT: Entergy Nuclear Northeast PLANT: Pilgrim Nuclear Power Station CALCULATION TITLE: Evaluation of Fluence Issues for PNPS Document Affected Project Manager Preparer(s) &

Revision Pages Revision Description Approval Checker(s)

Signature & Date Signatures & Date 1

3, 8, 9, 10, & 12 Update Sections 1.0, 4.5, T. J. Griesbach T. J. Griesbach 5.0, 6.0, and 7.0 5/11/07 5/10/07 G. L. Stevens 5/11/07 2

3, 6,7,9,8, 10, Update Sections 1.0, 4.3,

/1-12, 18 4.4, 5.0, 6.0, 7.0 and 4

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Table of Contents 1.0 IN TR O D U C T IO N................................................................................................................

3 2.0 TECHNICAL APPROACH............................................................

3 3.0 ASSUM PTIONS / DESIGN INPUTS...................................................................................

4 4.0 C A L C U L A T IO N S........................................................................................................................

4 5.0 RESULTS O F AN ALY SIS.....................................................................................................

9

6.0 CONCLUSION

S AND DISCUSSIONS................................................................................

10 7.0 R E F E R E N C E S...........................................................................................................................

11 List of Tables Table 1: Beltline Curve A for 54 EFPY with Bias Correction Factor on Fluence [8].................. 13 Table 2: PNPS ART Calculations for 54 EFPY with Bias Correction Factor on Fluence [9].....

14 Table 3: PNPS Charpy Upper Shelf Energy Values for 54 EFPY (Without 1.78 Bias Correction Factor on Fluence) [6]..................................................................................................

15 Table 4: PNPS Charpy Upper Shelf Energy Values for 54 EFPY (With 1.78 Bias Correction Factor on Fluence).......................................................................................................

16 Table 5: PNPS Maximum Projected Fluence and USE Drop for Vessel Beltline Materials......

17 Table 6: Effects of Irradiation on RPV Axial Weld Properties......................................................

18 Table 7: Effects of Irradiation on RPV Circumferential Weld Properties.....................................

19 List of Figures Figure 1: Pressure Test P-T Curve (Curve A) for 54 EFPY with Bias CF on Fluence [8]........... 20 Figure 2: Calculated Hydrotest Temperature and 1/4t ART versus Fluence.................................

21 Figure 3: Predicted Decrease in Upper Shelf Energy as a Function of Pct. Copper and Fluence...... 22 File No.: PNPS-27Q-301 Page 2 of 22 Revision: 2 Contains References to Proprietary Information F0306-OI RO

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1.0 INTRODUCTION

The recent fluence re-evaluation for the Pilgrim Nuclear Power Station (PNPS) reactor pressure vessel (RPV) using the EPRI RAMA code required an increased fluence bias correction factor (CF) of 1.78 to adjust for the benchmarking discrepancy with the cycle 4 surveillance capsule dosimetry results [ 1 ]. A rigorous technical explanation for this bias has not been determined. As a result, the NRC will not accept the PNPS fluence calculations for application to future plant operation without further justification. In response to this, Entergy Nuclear Northeast (ENN) has requested SI to perform an additional evaluation to demonstrate adequate vessel life prediction through the extended 60-year operating period with respect to the fluence projections.

Increasing fluence has an effect on the toughness of the RPV materials. This is measured by an increase in the adjusted reference temperature (ART) and a decrease in the upper shelf energy (USE) of the RPV beltline materials. The PNPS FSAR identifies the vessel as being controlling for all reactor pressure boundary carbon steel components [ 17]. The ASME Code [2] and 10CFR50, Appendix G [3] give criteria for maintaining pressure boundary integrity including the effects of materials degradation due to irradiation damage. Additional evaluations for equivalent margins have been submitted to the Nuclear Regulatory Commission (NRC) and approved for use by boiling water reactors (BWRs) for Charpy USE drop. These equivalent margin analyses for USE are published in BWRVIP-74-A [4]. In addition, BWRVIP-05 [5] provides a technical basis for alternative inspection requirements of the RPV shell welds to eliminate inspections of circumferential welds. BWRVIP-05 and the Supplemental SER [18] provide the basis for acceptable limits for BWR reactor vessel axial welds. The methods and criteria in these documents form the basis for demonstrating vessel integrity niargins, including the effects of plant aging due to fluence.

ENN performed an integrated plant assessment (IPE) to extend the operating license of PNPS.

This included a review of the time-limited aging analyses (TLAA) and exemptions to 10CFR50 for the period of extended operation [6]. Increasing fluence is one aspect considered in the TLAAs. The calculated fluence in the vessel using the method from Reference [1] is projected through 54 effective full power years (EFPY) without a bias correction factor. The results of that study are now being reevaluated using assumed fluences greater than the previously calculated results. This analysis for PNPS uses the established methods and criteria for evaluating embrittlement for fluence levels exceeding the previously projected end-of-license fluence in the vessel.

2.0 TECHNICAL APPROACH The shift in the ART and a decrease in the USE for ferritic materials are predicted by Regulatory Guide 1.99, Revision 2 [7]. The embrittlement trend curves are a function of File No.: PNPS-27Q-301 Page 3 of 22 Revision: 2 Contains References to Proprietary Information F0306-O1RO

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copper (Cu) content, nickel (Ni) content, and fluence; different trend curves apply for welds and base metals. The materials in the RPV that must be monitored for irradiation effects are the regions where significant fluence levels are projected (> lxl017 n/cm2, E > 1 MeV), and those materials are characterized as beltline materials. Analyses of all of the beltline materials for the PNPS vessel determine the weld or plate that is the limiting RPV beltline material. The properties of that limiting beltline material are then used to calculate the operating heatup, cooldown, and pressure test curves. Those calculations were performed previously for the PNPS RPV for a fluence up to 54 EFPY using fluence projections with and without the 1.78 bias correction factor [8, 9]. The calculations show that there is no RPV integrity concern for 54 EFPY even with the bias corrected fluence.

To further demonstrate that the fluence uncertainty issue for PNPS is not a concern, additional analyses are being performed in this calculation assuming even greater fluence levels in the RPV beyond the 54 EFPY predicted fluence values with the 1.78 bias correction factor.,The fluence levels are assumed to increase until a criterion for operability can no longer be maintained. When that limit is determined, the calculated factor on fluence is an indication of the conservatism against brittle fracture of the RPV (or some other criteria) in order to accommodate the observed uncertainty in the fluence calculations.

3.0 ASSUMPTIONS / DESIGN INPUTS

1. The pressure for the pressure test is normal operating pressure (1,035 psig) from Reference [10].

2. The maximum test temperature for the hydrotest is 212'F per the PNPS Technical Specifications

[11]. (Note that this is an operational limit, not a brittle fracture limit.)

4.0 CALCULATIONS 4.1 Maximum Fluence to Perform Hydrotest Irradiation by high energy neutrons raises the RTNDT of the reactor vessel materials. The ART is defined as RTNDT + ARTNDT + Margin in accordance with Regulatory Guide 1.99, Rev. 2 [7].

The pressure-temperature (P-T) curves are developed from the ART value for the vessel material. The calculated hydrotest pressure vs. temperature curve (Curve A) results for 54 EFPY are shown in Table 1 and in Figure 1 [8]. The PNPS projected values for ARTNDT and ART at 54 EFPY were calculated with the 1.78 bias correction factor on fluence [9]. The projected values of ART are shown in Table 2. The hydrotest pressure is the normal operating pressure, which is 1,035 psig [10]. The system hydrostatic test temperature is calculated to meet the requirements of ASME Section XI, Appendix G, Article G-2400 [2]. The system hydrostatic test should be performed at a temperature not lower than the highest required temperature for any component in the system. For PNPS, the limiting component is the beltline File No.: PNPS-27Q-301 Page 4 of 22 Revision: 2 Contains References to Proprietary Information F0306-OIRO

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material with the highest ART value at the quarter-thickness (1/4i) location. From Table 2, the limiting materials are the lower intermediate shell longitudinal welds #1 and #3.

The maximum calculated ART value for these welds at 54 EFPY is 122.7°F. This corresponds to a 1/4t fluence value of 1.46x10' 8 n/cm 2, including the 1.78 bias correction factor. The hydrotest temperature at this fluence is 152.5°F. This hydrotest temperature is interpolated linearly from the values from Table 1 as follows:

Hydrotest Temperature (0F) 150 152.5 155 Hydrotest Pressure (psig) 1,007 1,035 1,063 The temperature difference between the 1/4t ART value and the hydrotest temperature is calculated to be 29.8°F. This temperature difference is assumed to be constant for increasing fluence and ART values, so the maximum fluence to conduct the hydrotest can be calculated from the maximum achievable temperature to perform the hydrotest, which is 212'F for PNPS

[11]. The 1/4t fluence and corresponding 1/4t ART for the limiting welds are increased until the hydrotest temperature of 212'F is reached. From the table below it is noted that the maximum 1/4t fluence of 4.12x1018 n/cm 2 corresponds to a 1/4t ART value of 182.2°F for a hydrotest temperature of 212'F, the maximum temperature to perform the hydrotest at PNPS.

Calculation of Hydrotest Maximum Temperature and Fluence 1/4t Fluence (nlcmn) 1.46E+18 2.OOE+18 3.OOE+18 4.OOE+18 4.12E+18 4.50E+18 5.00E+18 1/4t ART (OF) 122.7 139.6 162.9 180.4 182.2 187.8 194.4 Hydrotest Temp.

(OF) 152.5 169.4 192.7 210.2 212 217.6 224.2 Temp.

Difference (OF) 29.8 29.8 29.8 29.8 29.8 29.8 29.8 Fluence Ratio 1.00 1.37 2.05 2.74 2.82 3.08 3.42 maximum fluence to conduct hydrotest < 212°F The calculated hydrotest temperature and 1/4t ART values versus fluence are shown in Figure 2.

A fluence ratio of 2.82 is the ratio of the maximum 1/4t fluence at the limiting vessel beltline welds compared to the 54 EFPY fluence with the 1.78 bias correction factor. In other words, the fluence with the 1.78 bias correction factor would have to be increased by an additional factor of 2.82 before the limiting hydrotest temperature of 212'F is reached.

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4.2 Maximum Fluence to Maintain Charpy Upper Shelf Energy Appendix G of 1 OCFR50 requires that reactor vessel beitline 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." Regulatory Guide 1.99, Rev. 2, Radiation Embrittlement of Reactor Vessel Materials, defines the method for predicting upper shelf energy drop in terms of a percentage from the unirradiated value. Figure 3 shows the predicted Charpy upper shelf energy for welds and base metals as a function of copper content and fluence.

The predicted Charpy upper shelf energy (CvUSE) values for PNPS at 54 EFPY were determined previously for the PNPS license renewal project [6]. The predicted CRUSE values based on the Regulatory Guide 1.99 Position 1 method are shown in Table 3. The predicted values for CvUSE using the 54 EFPY fluences with the 1.78 bias correction factor are shown in Table 4. It isnoted that all projected USE values are above 50 ft-lbs, even with the 1.78 bias correction factor on fluence. The USE limit shows a minimum fluence ratio of 4.9 for the projected fluence to reach 50 ft-lbs for the lower intermediate shell axial welds, as shown in Table 5. Because the USE values are always greater than 50 ft-lbs., the equivalent margin method of BWRVIP-74-A is not required.

4.3 Maximum Fluence Bounded by the Reactor Vessel Weld Failure Probability The BWVIP recommendations for inspection of reactor vessel shell welds in BWRVIP-05 [5]

are based on generic analyses supporting a Safety Evaluation Review (SER) conclusion that the generic plant axial weld failure rate is no greater than 5x10-6 per reactor year [12] at the end of 40 years. 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 tojustify 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 [ 13]. 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 were less than the values calculated in the BWRVIP-05 SER [12].

Table 6 contains a comparison of the PNPS reactor vessel limiting axial weld parameters to those used in the NRC analysis. The data in column two is from the NRC Supplemental SER on the BWRVIP-05 Report, and it is the basis for evaluating axial welds in BWRs [18]. The data in the third column (PNPS) is the projected 54 EFPY data for PNPS without the 1.78 bias correction factor on fluence [6]. (For consistency with the NRC evaluation, the RTNDT is calculated without the margin term.) The data in column four (PNPS with Bias CF) is the projected 54 EFPY data for PNPS with the 1.78 bias correction factor on fluence. Column File No.: PNPS-27Q-301 Page 6 of 22 Revision: 2 Contains References to Proprietary Information F0306-O1 RO

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five (PNPS Limit) shows the maximum fluence and RTNDT to assure that the limiting axial weld remains bounded by the mean value of RTNDT < I I4°F from the NRC Supplemental SER

[18]. The mean RTNDT limit of 114'F was chosen to represent a vessel failure frequency due to failure of the axial welds of less than 5x 10-0 per reactor-year. Although this analysis was performed for the initial 40-year license period, it was considered to be applicable for the license renewal period per the guidance in the Supplemental SER. The maximum ID fluence of 3.37x1018 n/cm 2 gives a fluence ratio of 1.66 compared to the 54 EFPY fluence with the 1.78 bias correction factor.

Table 7 contains a comparison of the PNPS reactor vessel limiting circumferential weld parameters to those used in the NRC analysis. The data in column two (CE) is from Table 2.6-5 ofthe, NRC SER for BWRVIP-05 [12]. The data in the third column (PNPS) is the projected 54 EFPY data for the PNPS circumferential weld without the 1.78 bias correction factor on fluence [6]. The data in column four (PNPS with Bias CF) is the projected 54 EFPY data for the PNPS circumferential weld with the 1.78 bias correction factor on' fluence. Column five (PNPS Limit) shows the maximum fluence and RTNDT to assure that the PNPS circumferential weld remains bounded by the value of 128.5'F determined from the CEOG and accepted in the SER [12]. The maximum ID fluence of 1. 14x10' 9 n/cm 2 gives a fluence ratio of 7.35 compared to the 54 EFPY fluence with the 1.78 bias correction factor.

PNPS obtained relief from the examination of RPV circumferential welds related to the augmented shell weld examination requirements contained in 1 OCFR50.55a(g)(6)(ii)(A)(5).

The reduction in scope of these inspections from essentially 100 percent of all RPV shell welds to examination of essentially 100 percent of the axial welds and essentially zero percent of the circumferential welds was based on the NRC staff determination that the conditional probability of failure for these welds was within the acceptable limits at the expiration of the current operating license [13]. The results in Tables 6 and 7 show that the bounding reactor vessel weld, conditional failure probabilities can be maintained well beyond the 54 EFPY projected fluences and ART values for the PNPS vessel. The large calculated fluence ratio shown in Table 7 indicates that the criteria for circumferential vessel welds will not be the limiting factor for fluence margin in the PNPS RPV. However, the fluence ratio of 1.66 shown in Table 6 to assure that the limiting axial weld RTNDT (and equivalent failure frequency) is acceptable makes the axial welds in the PNPS vessel the limiting concern with respect to fluence.

4.4 Effect of Fluence on Evaluation of N2 Nozzles The fluence levels in the N2 nozzles are relatively low compared to the peak fluence in the beltline. These fluences shown in the table below were obtained from the RAMA code fluence calculation [9, 14].

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54 EFPY Fluence @ 114t 54 EFPY Fluence @ 1/4t (wlo 1.78 bias CF)

(with 1.78 bias CF)

(n/cmA2)

(n/cmA2)

Recirc. inlet (N2) nozzles 2.02E+17 3160E+17 Limiting Axial Welds 8.18E+17 1.46E+1 8 The effect of the increasing fluence on the calculated ART values for the limiting weld and the N2 nozzles is shown below. The ART values for the A508-2 nozzle forgings was estimated using upper bound Cu = 0.35, Ni = 0.85, and an initial RTNDT of 0°F [14].

54 EFPY ART @ 114t 54 EFPY ART @ 1/4t (w/o 1.78 bias CF)

(with 1.78 bias CF)

(0F)

(OF)

Recirc. inlet (N2) nozzles 77.0 94.7 Limiting Axial Welds 95.3 122.7" Structural Integrity Associates recently performed an evaluation of the recirc. inlet nozzles using best estimate copper and nickel chemistry values of Cu = 0.15 wt%, Ni = 0.85 wt% [14].

Using these best estimate values, the calculated ART values for the nozzles are as follows:

54 EFPY ART @ 1/4t 54 EFPY ART @ 1/4t (w/o 1.78 bias CF)

(with 1.78 bias CF)

_(*F)

(*F)

Recirc. inlet (N2) nozzles 39.9 56.4 Limiting Axial Welds 95.3 122.7*

From the comparison of the ART values for the recirc. inlet nozzles and the limiting axial welds, the recirc. inlet nozzle embrittlement levels are well below the projected ART values for the limiting axial welds. This is mainly because of significantly lower fluences at the height of the nozzles compared to the active core region. Thus, there is no impact of fluence uncertainty for this evaluation, and it is determined that the nozzles will not become the limiting beltline materials for P-T limits or hydrotest conditions as fluence levels are increased.

• Note: The limiting axial weld values for ART = RTNDT + ARTNDT + Margin Term (see Table 2);

these values are different than the calculated mean RTNDT values for the limiting axial welds shown in Table 6 that do not include the Margin Term.

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4.5 Effect of Fluence on RPV Internals 4.5.1 Top Guide BWRVIP-26 calculated the minimum top guide fluence for 32 EFPY (40 years) as 4x10 2" n/cm2 [15].' The threshold for IASCC is 5x10 20 n/cm2, and the PNPS top guide fluence will exceed this threshold [6]. Therefore, PNPS must manage IASCC of the top guide assembly.

PNPS has implemented the inspection recommendation in BWRVIP-26 through the BWR Vessel Internals Program [16]. The BWR Vessel Internals Program will adequately manage the effects of aging on the top guide for the period of extended operation. 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.

4.5.2 Corpe Shroud The core shroud is a BWR component that is known to be susceptible to aging effects. Section 3.8.12 of the PNPS License Renewal Project, TLAA and Exemption Evaluations [6] addresses the time limited aging analyses of the core shroud. A review of the analyses related to the core shroud found that the only TLAA involves the fatigue analysis and calculation of cumulative usage factors (CU~s) for the shroud repair. The core shroud does not 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.

5.0 RESULTS OF ANALYSIS The effects of increased fluence beyond the projected 54 EFPY fluence calculations for the PNPS RPV are summarized below for each of the potential aging effects. The results are compared to determine the minimum acceptable fluence ratio. This is the fluence multiplier that could be achieved compared to the 54 EFPY fluence with the 1.78 bias correction factor, and is the measure of tolerance on fluence before a limit is reached that would exceed a Code limit, regulatory criterion, or service limit.

Effect of Fluence on Acceptable Fluence Ratio Hydrotest Temperature 2.82 Charpy Upper Shelf Energy 4.86 RPV Axial Weld Failure Probability 1.66*

RPV Circ. Weld Failure Probability 7.35 Evaluation of N2 Nozzles Bounded by beltline

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6.0 CONCLUSION

S AND DISCUSSIONS Fluence contributes to changes in the vessel beltline material properties. These changes are measured by' the shift in RTNDT or the drop in USE of the ferritic. materials (i.e., welds, plates, and forgings). The analyses using projected fluence values for license renewal (54.EFPY) for PNPS show no limitations due to embrittlement concerns for the vessel. Considering increasing fluence levels, the RPV analyses demonstrate that the Code and regulatory criteria can be met for operation well beyond this maximum fluence level by a factor of 1.66 (or greater) on the 54 EFPY fluence including a bias correction factor of 1.78.

The limiting condition for the PNPS vessel due to fluence is the maximum level of embrittlement of the axial welds of 114'F to remain below a calculated reactor vessel failure frequency of 5x 10.6 per reactor-year.

The next limiting condition for the vessel is the temperature required to perform the ASME Code hydrotest. The temperature to perform the hydrotest is prescribed by ASME Section XI, Article G-2400 that requires a safety factor of 1.5 on the pressure stress intensity to prevent brittle fracttire of the vessel during this test. The maximum temperature limit for the hydrotest of 212'F in the PNPS Technical Specifications is an administrative limit; it may be possible to perform the test at higher temperatures which would allow for even higher fluence. levels.

These analyses demonstrate that there is a considerable tolerance on the acceptable range of fluence. This is exemplified by the difference between the fluence for the maximum predicted levels of embrittlement and the limiting criteria for axial weld failure frequency, a difference large enough to accommodate the uncertainties on the calculated fluence for PNPS.

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7.0 REFERENCES

1.

ENT-FLU-001-R-001, "Pilgrim Nuclear Power Station Reactor Pressure Vessel Fluence Evaluation at End of Cycle 15 and 54 EFPY," TransWare Enterprises Inc.,

December 2005, Proprietary, (SI File No. PNPS-22Q-201).

2.

ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, 1998 Edition, 2000 Addenda.

3.

Code of Federal Regulations, 10 CFR, Part 50, Appendix G, Federal Register: January 2005.

4.

BWRVIP-74-A, "BWR Reactor Vessel Inspection and Flaw Evaluation Guidelines for License Renewal, BWR Vessel and Internals Project," EPRI Proprietary, 1008872, June 2003, (SI File No. BWRVIP-01-274AP).

5.

BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," BWR Vessel and Internals Project," EPRI Proprietary, EPRI TR-105697, September 1995. (SI File No. BWRVIP-01-205P).

6.

PNPS License Renewal Project, TLAA and Exemption Evaluations, LRPD-03, Revision 0, January 2006, (SI File No. PNPS-27Q-202).

7.

NRC Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

8.

SI Calculation PNPS-03Q-301, Revision 1, "Development of Pressure Test (Curve A)

P-T Curves," 1/30/06.

9.

SI Calculation PNPS-22Q-301, "ARTNDT and ART Evaluation," 1/20/06.

10.

Email from Bryan Ford to Raymond Pace, Timothy J. Griesbach, and Gary L. Stevens,

Subject:

Maximum Pressure Test Temperature, dated 3/8/07, (SI File No. PNPS-27Q-205).

11.

PNPS Technical Specifications, Revision 274, Amendment No.'s 224 and 225, Limiting Conditions for Operation, 3.14 Special Operations, A. Inservice Hydrostatic and Leak Testing Operation, (SI File No. PNPS-27Q-206).

12.

BWRVIP-05 SER (Final), USNRC letter from Gus C. Lainas to Carl Terry, Niagara Mohawk Power Company, BWRVIP Chairman, Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report, (TAC No. M93925), July 28, 1998, (SI File No. BWRVIP-01-205P).

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13.

Letter from J. Boska (NRC) to M. Bellamy (ENGC), Pilgrim Nuclear Power Station -

Pilgrim Relief Request No. 28, Relief from ASME Code,Section XI, Examinations of Reactor Pressure Vessel Circumferential Shell Welds (TAC No. MB6074), April 11, 2003, (SI File No. PNPS-27Q-208).

14.

SI Calculation PNPS-22Q-302, "N2 Nozzle Evaluation," 2/21/06.

15.

BWRVIP-26, "BWR Top Guide Inspection and Flaw Evaluation Guidelines (BWRVIP-26)," EPRI Proprietary, EPRI Report TR-107285, December 1996, (SI File No.

BWRVIP-01-226P).

16.

Engineering Report PNPS-EP-06-0000 1, Revision 0, "Reactor Vessel Internals Inspection Program," (SI File No. PNPS-27Q-207)

17.

Pilgrim Nuclear Power Station Final Safety Analysis Report, Section 3.3, Reactor Vessel Internals Mechanical Design, and Section 4.2, Reactor Vessel and Appurtenances Mechanical Design, (SI File No. PNPS-27Q-209).

18.

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Supplemental Safety Evaluation of EPRI Topical Report TR-105697 "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)", enclosure from letter to Mr. Carl Terry (BWRVIP Chairman, NMPC) from Mr. Jack Strosnider (Director, NRR), March 7, 2000, (SI File No. PNPS-27Q-212).

File No.: PNPS-27Q-301 Revision: 2 Page 12 of 22 Contains References to Proprietary Information F0306-01RO

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Table 1: Beltline Curve A for 54 EFPY with Bias Correction Factor on Fluence [81 Pressure-Tem pera ture Curve Calcula tion (Pressure Test = Curve A)

(NOTE: THE ARTNDT includes a calculated bias on fluence of 1.78.)

Plant =

Component =

Vessel thickness, t =

Vessel Radius, R =

ARTNoT =

Cooldown Rate, CR =

KIT =

AT,/4t =

Safety Factor =

Mr Temperature Adjustment =

Height of Water for a Full Vessel =

Pressure Adjustment Hydro Test Pressure =

Flange RTNOT =

inches, so

=

2.352

-,inch inches oF======>

b,$

5 41'..

°F/hr ksi'inch" 2 (From Appendix G, for cooldown rate above)

• F (no thermal for pressure test)

(for pressure test)

(From Appendix G, for inside surface axial flaw) oF inches psig (hydrostatic pressure for a full vessel at 70°F)

Fluid Temperature T

(°F) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 170.0 175.0 180.0 185.0 190.0 195.0 1114t Temperature

(°F) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 170.0 175.0 180.0 185.0 190.0 195&0 Kic (ksi*inch1l2 )

40.43 40.43 41.19 42.03 42.95 43.98 45.11 46.37 47.75 49.28 50.97 52.84 54.91 57.19 59.72 62.51 65.59 68.99 72.76 76.92 81.52 86.60 92.21 98.42 105.28 112.86 121.24 Kip (ksi*inchlI2 )

26.95 26.95 27.46 28.02 28.64 29.32 30.08 30.91 31.84 32.86 33.98 35.23 36.61 38.13 39.81 41.67 43.73 46.00 48.51 51.28 54.34 57.73 61.48 65.61 70.19 75.24 80.83 psig

-F Calculated Pressure P

(psig) 0 601 612 625 638 654 671 689 710 733 758 785 816 850 888 929 975 1026 1082 1143 1212 1287 1371 1463 1565 1678 1802 Adjusted Temperature for P-T Curve (IF) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 170.0 175.0 180.0 185.0 190.0 195.0 Adjusted Pressure for P-T Curve (psig) 0 583 594 606 620 635 652 671 691 714 739 767 798 832 869 911 957 1,007 1,063 1,125 1,193 1,269 1,352 1,445 1,547 1,659 1,784 File No.: PNPS-27Q-301 Revision: 2 Page 13 of 22 Contains References to Proprietary Information F0306-O1RO

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)

Table 2: PNPS ART Calculations for 54 EFPY with Bias Correction Factor on Fluence 191 Pilgrim RPV Material ART Calculations (NOTE: This table covers all RPV materals wth an exposed fluence, E > 1 MeV. of greater than l.Ox1O" nlcm 2.)

(54 EFPY) includes 1 78 calculated bias on fluence M

Estimated Chemistry Chemistry Ad]ustments For 1/4t Description Piece Code Heat Initial RT.

Factor ARTNOT Margin Terms ART T No.

No.

No.

(IF)

Cu(wt%) NI(wt%)

(=F)

(=F) oArF) o(°F)I

(-F)

LowerShell#1 337-01A G-3109-2 C-2957-2 0

0.10 0.47 65.0 30.5 15.3 0.0 61.1 LowerShell#2 337-01B G-3109-1 C-2957-1

-3 0.10 0.48 65.0 30.5 15.3 0.0 58.1 LowerShell#3..

337-LC '. G-3109-3

'C-2973-1

-4 0.11 0.63 74.5 35.0 17.0 0.0 65.0 Lower-int. Shell #1 337-03A G-3108-3 C-2945-2

-12 0.10 0.66 65.6 34.3 17.0 0.0 56.3 Lower-int. Shell #2 337-03B G-3108-1 C-2921-2

-30 0.14 0.60 100.0 52.2 17.0 0.0 56.2 Lower-int. Shell #3 337-03C G-3108-2 C-2945-1

-7 0.10 0.65 65.5 34.2 17-0 0.0 61.2 Ha Estimated Chemistry Chemistry Ad!ustments For 114t Description Seam Heat Flux Type &

Initial RTtMT Factor ARTNDT Margin Terms ARTmrN No.

No.

Lot No.

(IF)

Cu (wt%)l NI(wt%)

(°F)

(=F) aa (*F) ai(°F)

(°F)

L. Int Shell Long. Weld #l

1-338A 27204112008 Lindea1092 3774

.48 0.219..

0.996.

231,1 114.7 28.0 0.0 122.7 L. kI. Shell Long. Weld #2 1-338B.

27204112008 Linde 1092 3774

-48 0.219 0.996 231.1 78.4 28.0 0.0 86.4 L. Int. Shell Long. Weld #3 "1338Ci. 27204112008 Llnde1092 #3774

-48 0.219 0.996 231.1 114.7 28.0 0.0 122.7 L. Int.IL. S hell Girth Wedl 1-344 21935 Linde 1092 #3869

-50 0.183 0.704 172.2 75.4 28.0 0.0 81.4 Lower Shell Long. Weld #1 2-338A 27204 Linde 1092 #3714

-34 0.203 1.018 226.8 83.5 28.0 0.0 105.5 Lower Shell Long. Weld #2 2-338B:

27204 Linde 1092 #3714

-34 0.203 1.018 226.8 96.5 28.0 0.0 118.5 Lower Shell Loro. Weld #3 2,338C:

27204 Linde 1092 #3714

-34 0.203 1.018 226.8 87.8 28.0 0.0 109.8 Fluence Information (see Note.2):

Calculated Fluence Bias =

1.78 Wall Thickness (inches)

Fluence at ID Attenuation, 114t Fluence @ 1/4t Fluence Factor, FF Locatio n Full (3) 1 /4t (n/cm 2 )

e "°'u (n /cm

2) f0.2 o I.0 ) )

Lower Shell#1 5.531 1.383 1.80E+18 0.718 1.29E+18 0.470 Lower Shell #2 5.531 1.383 1.80E+18 0.718 1.29E+18 0.470 Lower Shell#3 5.531 1.383 1.80E+18 0.718 1.29E+18 0.470 Lower-knt Shell #1 5.531 1.383 2.28E+18 0.718 1.63E+18 0.522 Lower-int. Shell #2 5.531 1.383 2.28E+18 0.718 1.63E+18 0.522 Lower-int. Shell #3 5.531 1.383 2.28E+18 0.718 1.63E+18 0.522 L-inL Shell Long. Weld #1 5.531 1.383 2.03E+18 0.718 1.46E+18 0.496 L. Int. Shell Long. Weld #2 5.531 1.383 9.20E+17 0.718 6.60E+17 0.339 L. int. Shell Long. Weld #3 5.531 1.383 2.03E+18 0.718 1.46E+18 0.496 L. intL... Shell Girth Weld 5.531 1.383 1.55E+18 0.718 1.11E+18 0.438 Lower Shell Long. Weld #1 5.531 1.383 1.08E418 0.718 7.77E+17 0.368 Lower Shell Long. Weld #2 5.531 1.383 1.45E+18 0.718 1.04E+18 0.425 Lower Shell Long. Weld #3 5.531 1.383 1.20E+18 0.718 8.58E+17 0.387 Notes:

1. Material information taken from SIA Report No. SIR-00-082. Revision 0, "Updated Evaluation of Reactor Pressure Vessel Materials Properties for Pilgrim Nuclear Power Station," August 2000. Tables 3-1 through 3-12.

2. Fluence values from Transware Report No. ENT-FLU-001-R-001, Revision 0. "Pilgrim Nuclear Power Station Reactor Pressure Vessel Fluence Evaluation," Tables 7-3 and 7-4, and are multiplied by a calculated bias of 1.78.

3. RPV minimum thickness = 5 17132" per Section 3.3.2 of SIR-00-082, Revision 0.

I.

File No.: PNPS-27Q-301 Revision: 2 Page 14 of 22 Contains References to Proprietary Information F0306-O1RO

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Table 3: PNPS Charpy Upper Shelf Energy Values for 54 EFPY (Without 1.78 Bias Correction Factor on Fluence) [61 Material Description 54 EFPY Projection 114 T fluenc

% Drop in Reactor Vessel Material Unirradiated 014 n/cm)

USE USE (1%4 T")

Bettline Region Location Maoo Type Identification Heat #

%Cu CvUSE Lower Intermediate Shell A5338 G-3108-1 C-2921-2 0.14 81 0.084 12.79%

70.6 Lower Intermediate Shell A533B G-3108-2 C-2945-1 0.10 80 0.084 10.57%

71.5 Lower Intermediate Shell A533B G-3108-3 C-2945-2 0.10 81 0.084 10.57%

72.4 Lower Shell A5335 C-3109-1 C-2957-1 0.10 76 0.061 9.79%

68.6 Lower Shell A533B G-3109-2 C-2957-2 0.10 79 0.061 9.79%

71.3 Lower Shell A533B G-3109-3 C-2973-1 0.11 72 0.061 10.31%

64.6 Lower Int/Lower Shell Circ Weld Linde 1092 1-334 21935 0.18 75 0.057 16.39%

62.7 Lower Int Shell Axial Welds Unde 1092 1-338AM,,C 27204-12008 0.22 75 0.076 19.52%

60.4 Lower Shell Axial Welds Linde 1092 2-338AB,C 27204 0.20 75 0.050 16.87%

62.3 File No.: PNPS-27Q-301 Revision: 2 Page 15 of 22 Contains References to Proprietary Information F0306-OIRO

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Table 4: PNPS Charpy Upper Shelf Energy Values for 54 EFPY (With 1.78 Bias Correction Factor on Fluence)

Material Description 54 EFPY Projection (with 1.78 bias CF on fluence)

Reactor Vessel Matl Matl Heat#

%Cu Unirr.

1/4t fluence

% Drop USE Beltline Region Location Type Ident.

CvUSE (10A19 n/cm2) in USE

@ 1/4t Lower Intermediate Shell A533B G-3108-1 C-2921-2 0.14 81 0.129 14.3 69.4 Lower Intermediate Shell A533B G-3108-2 C-2945-1 0.10 80 0.129 11.7 70.6 Lower Intermediate Shell A533B G-3108-3 C-2945-2 0.10 81 0.129 11.7 71.5 Lower Shell A533B G-3109-1 C-2957-1 0.10 76 0.163 12.3 66.7 Lower Shell A533B G-3109-2 C-2957-2 0.10 79 0.163 12.3 69.3 Lower Shell A533B G-3109-3 C-2973-1 0.11 72 0.163 13.1 62.6 Lower Int./Lower Shell Circ. Weld Linde 1092 1-334 21935 0.183 75 0.111 19.6 60.3 Lower In. Shell Axial Welds Linde 1092 1-338A,B,C 27204/12008 0.219 75 0.146 23.2 57.6 Lower Shell Axial Welds Linde 1092 2-338A,B,C 27204 0.203 75 0.104 20.5 59.6 File No.: PNPS-27Q-301 Revision: 2 Page 16 of 22 Contains References to Proprietary Information F0306-0 1 RO

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Table 5: PNPS Maximum Projected Fluence and USE Drop for Vessel Beltline.Materials Material Description Maximum Projected Fluence and USE Drop Reactor Vessel MatU Matl

%Cu Unirr.

1/4t fluence Max. % Drop Min. USE Fluence Beltline Region Location Type Ident.

CvUSE (10A19 n/cm2) in USE

@ 1/4t Ratio Lower Intermediate Shell A533B G-3108-1 0.14 81

> 6.0 38.3 50.0

> 46.5 Lower Intermediate Shell A533B G-3108-2 0.10 80

> 6.0 37.5 50.0

> 46.5

• Lower Intermediate Shell A533B G-3108-3 0.10 81

> 6.0 38.3 50.0

> 46.5 Lower Shell A533B G-3109-1 0.10 76

> 6.0 34.2 50.0

> 36.8 Lower Shell A533B G-3109-2 0.10 79

> 6.0 36.7 50.0

> 36.8 Lower Shell A533B G-3109-3 0.11 72

> 6.0 30.6 50.0

> 36.8 Lower Int./Lower Shell Circ. Weld Linde 1092 1-334 0.183 75 1.11 33.3 50.0 10 Lower Int. Shell Axial Welds Linde 1092 1-338A,B,C 0.219 75 0.71 33.3 50.0 4.86*

Lower Shell Axial Welds Linde 1092 2-338A,B,C 0.203 75 0.86 33.3 50.0 8.3

• limiting fluence ratio to reach 50 ft-lbs CvUSE = (0.7 1 EI 9)/(0.146E 19) = 4.86 File No.: PNPS-27Q-301 Revision: 2 Page 17 of 22 Contains References to Proprietary Information F0306-0IRO

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Table 6: Effects of Irradiation on RPV Axial Weld Properties Limiting Axial Welds - Lower Int. Long. Welds #1 and #3 Wire Heat/Lot (27204/12008, Lot No. 3774)

PNPS with PP ii Plant PNPS Mod 2**

PNPS Bias CF PNPS Limit Parameter Description USNRC Data for Data for' Data for Limiting Plant-axial weld axial weld axial weld Specific Data (no bias CF)

(1.78 bias CF)

(limiting fluence)

EFPY 32**

54 54

>54 Initial (unirradiated) reference

-2**

-48

-48

-48 temperature (RTndt), _F Neutron fluence at the end of the requested relief period (Peak 1.14E+18 2.03E+18 3.37E+18*

Surface Fluence in the Beltline),

n/cmA2 FF = Fluence factor (calculated 0.444 0.573 0.701 per Reg. Guide 1.99, Rev. 2)

Weld Copper content, wt. %

0.219 0.219 0.219 Weld Nickel content, wt%

0.996 0.996 0.996 CF = Chemistry Factor 231.1 231.1 231.1 Increase in reference temperature 102.9 132.4 162.0 (ARTndt), 'F

(= FF*CF)

Mean adjusted reference 114.0"*

54.9 84.4 114.0 temperature (ART), °F

(= RTndt +ARTndt)

• Fluence ratio = (3.37E18)/(2.03E18) = 1.66

• • This is a variant of the PNPS limiting weld input data with initial RTNDT = -2°F to match a vessel failure frequency of the axial welds below 5x 10-6 per reactor-year for each BWR File No.: PNPS-27Q-301 Revision: 2 Page 18 of 22 Contains References to Proprietary Information F0306-0IRO

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Table 7: Effects of Irradiation on RPV Circumferential Weld Properties Limiting Circ. Weld - Lower Int.-to-Lower Shell Circ. Weld 1-344 Wire Heat/Lot (21935, Lot No. 3869)

Plant CE (CEOG)

PNPS PNPS with PNPS Limit Bias CF Parameter Description USNRC Data for Data for Data for Limiting Plant-circ. weld circ. weld circ. weld Specific Data (no bias CF)

(1.78 bias CF)

(limiting fluence)

EFPY 64 54 54

>54 Initial (unirradiated) reference 0

-50

-50

-50 temperature (RTndt), OF Neutron fluence at the end of the requested relief period (Peak 4.OOE+18 8.69E+17 1.55E+18 1.14E+19*

Surface Fluence in the Beltline),

n/cmA2 FF = Fluence factor (calculated 0.746 0.389 0.510 1.037 per Reg. Guide 1.99, Rev. 2)

Weld Copper content, wt. %

0.183 0.183 0.183 0.183 Weld Nickel content, wt%

0.704 0.704 0.704 0.704 CF = Chemistry Factor 172.2 1.72.2 172.2 172.2 Increase in reference temperature 128.5 67.1 87.9 178.5 (ARTndt), 'F

(=,FF*CF)

Mean adjusted reference 128.5 17.1 37.9 128.5 temperature (ART), -F

(=RTndt +ARTndt)

• Fluence ratio = (1.14E19)/(1.55E18) = 7.35 File No.: PNPS-27Q-301 Revision: 2 Page 19 of 22 Contains References to Proprietary Information F0306-01 RO

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Figure 1: Pressure Test P-T Curve (Curve A) for 54 EFPY with Bias CF on Fluence [81 1,200 1,100 1,000

-t r-j T -4 L

§211.11

.1:-i F

I ~i-1~1T7 --

~1~~

'a C.

0 lU a-co 0I.-

-.J LU Ci)

LU 0

C.)

tw 2

I.-

LU LU.

a..

900 800 700 600 500 400 300 200

-I 11

~'-~

~-~-i T7 I

I I

t -- r

.4 iI ITtXVp K-Kf r I

Li

'.1 Lt ZFIdZH

--K 7Z zz 1~~~~*

Ii 5~7 45 m

T-4i-zlzi

--- I----zL ii:::

-- I1 I

I I.

.LIII II,-

---

K--

I I~

r

-I.}I7 I

-Beltline 7--7-1 i

100 0

Bottom Head Upper Vessel

[-

VP-I

~-~--HIiZ~

0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

PNPS Pressure Test Curve (Curve A), 54 EFPY (NOTE: The fluence used on the beitline curve is increased by a calculated bias on fluence of 1.78.)

File No.: PNPS-27Q-301 Revision: 2 Page 20 of 22 Contains References to Proprietary Information F0306-O IRO

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Figure 2: Calculated Hydrotest Temperature and 1/4t ART versus Fluence 250 200 150 LL 0

E 50 0

0 1E+18 2E+18 3E+18 4E+18 5E+18 114t Fluence (n/cmA2) 6E+18 File No.: PNPS-27Q-301 Revision: 2 Page 21 of 22 Contains References to Proprietary Information F0306-0I RO

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Figure 3: Predicted Decrease in Upper Shelf Energy as a Function of Pct. Copper and Fluence (from Reg. Guide 1.99, Rev. 2 [71) 100 8

7 6

5 4

3 E

0 (3

0 0~

0 Cw 0

C,,

C 0(0w 0

U 0a 2

10 8

7 6

5.

4 3

2 1.E-I.OE+17 1.OE+18 1.OE+19 FLUENCE; n/cm 2 (E > 1MeV)

I.OE+20 File No.: PNPS-27Q-301 Revision: 2 Page 22 of 22 Contains References to Proprietary Information F0306-OIRO