ML070230286

License Renewal Application Amendment 12

ML070230286
Person / Time
Site:	Pilgrim
Issue date:	01/16/2007
From:	Bethay S Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.07.005, TAC MC9669
Download: ML070230286 (22)

Text

Entergy Nuclear Operations, Inc.En tr~gyPilgrim Station'cýý nteW600 Rocky Hill Road Plymouth, MA 02360 Stephen J. Bethay Director, Nuclear Assessment January 16, 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 12

REFERENCE:

Entergy letter, License Renewal Application, dated January 25, 2006 (2.06.003)

LETTER NUMBER: 2.07.005

Dear Sir or Madam:

In the referenced letter, Entergy Nuclear Operations, Inc. applied for renewal of the Pilgrim Station operating license. NRC TAC NO. MC9669 was assigned to the application.

This License Renewal Application (LRA) amendment consists of five attachments stemming from discussions with the NRC license renewal staff. Attachment A contains the list of revised regulatory commitments.

Attachment B contains responses to requests for information conveyed in NRC letter dated December 29, 2006. Attachment C contains a supplemental response to request for additional information on RAI 4.3.1.2-2 contained in LRA Amendment 9.Attachment D contains the response to a supplemental request for supplemental information on environmentally-assisted fatigue. Attachment E contains the response to request for a commitment to groundwater sampling.Please contact Mr. Bryan Ford, (508) 830-8403, if you have any questions regarding this subject.I declare under penalty of perjury that the foregoing is true and correct. Executed on January Io, 2007. Ste hnJIt~y Director, Nuclear Safety Assessment DWE/dI Attachments: (as stated)cc: see next page Entergy Nuclear Operations, Inc.Pilgrim Nuclear Power Station Letter Number: 2.07.005 Page 2 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 Shea Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 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 Schrafft Center, Suite 1 M2A 529 Main Street Charlestown, MA 02129 Ms. Cristine McCombs, Director Massachusetts Emergency Management Agency 400 Worchester Road Framingham, MA 01702 Mr. James E. Dyer, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-00001 ATTACHMENT A to Letter 2.07.005 (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 No./Comments 1 Implement the Buried Piping and Tanks Inspection June 8, 2012 Letters B.1.2/Audit Program as described in LRA Section B.1.2. 2.06.003 Item 320 and 2.06.057 2 Enhance the implementing procedure for ASME June 8, 2012 Letters B.1.6/Audit Section XI inservice inspection and testing to specify 2.06.003 Item 320 that the guidelines in Generic Letter 88-01 or and approved BWRVIP-75 shall be considered in 2.06.057 determining sample expansion if indications are found in Generic Letter 88-01 welds.3 Inspect fifteen (15) percent of the top guide locations As stated in the Letters B.1.8 / Audit using enhanced visual inspection technique, EVT-1, commitment 2.06.003 Items 155, within the first 18 years of the period of extended and 320 operation, with at least one-third of the inspections to 2.06.057 be completed within the first six (6) years and at least and two-thirds within the first 12 years of the period of 2.06.064 extended operations.

Locations selected for and examination will be areas that have exceeded the 2.06.081 neutron fluence threshold.

4 Enhance the Diesel Fuel Monitoring Program to June 8, 2012 Letters B.1.10/include quarterly sampling of the security diesel 2.06.003 Audit Items generator fuel storage tank. Particulates (filterable and 320, 566 solids), water and sediment checks will be performed 2.06.057 on the samples. Filterable solids acceptance criteria and will be = 10 mg/l. Water and sediment acceptance 2.06.089 criteria will be = 0.05%.5 Enhance the Diesel Fuel Monitoring Program to June 8, 2012 Letters B.1.10 /install instrumentation to monitor for leakage between 2.06.003 Audit Items the two walls of the security diesel generator fuel and 155, 320 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).

1 2.06.0571 I COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section No./Comments 7 Enhance Fire Protection Program procedures to state June 8, 2012 Letters B.1.13.1 /that the diesel engine sub-systems (including the fuel 2.06.003 Audit Items supply line) shall be observed while the pump is and 320, 378 running. Acceptance criteria will be enhanced to 2.06.057 verify that the diesel engine did not exhibit signs of and degradation while it was running; such as fuel oil, 2.06.064 lube oil, coolant, or exhaust gas leakage. Also, enhance procedures to clarify that the diesel-driven fire pump engine is inspected for evidence of corrosion in the intake air, turbocharger, and jacket water system components as well as lube oil cooler.The jacket water heat exchanger is inspected for evidence of corrosion or buildup to manage loss of material and fouling on the tubes. Also, the engine exhaust piping and silencer are inspected for evidence of internal corrosion or cracking.8 Enhance the Fire Protection Program procedure for June 8, 2012 Letters B.1.13.1 /Halon system functional testing to state that the 2.06.003 Audit Item Halon 1301 flex hoses shall be replaced if leakage and 320 occurs during the system functional test. 2.06.057 9 Enhance Fire Water System Program procedures to June 8, 2012 Letters B.1.13.2/include inspection of hose reels for corrosion.

2.06.003 Audit Item Acceptance criteria will be enhanced to verify no and 320 significant corrosion.

2.06.057 10 Enhance the Fire Water System Program to state that June 8, 2012 Letters B.1.13.2 /a sample of sprinkler heads will be inspected using 2.06.003 Audit Item guidance of NFPA 25 (2002 Edition) Section and 320 5.3.1.1.1.

NFPA 25 also contains guidance to repeat 2.06.057 this sampling every 10 years after initial field service testing.11 Enhance the Fire Water System Program to state that June 8, 2012 Letters B.1.13.2 /wall thickness evaluations of fire protection piping will 2.06.003 Audit Item be performed on system components using non- and 320 intrusive techniques (e.g., volumetric testing) to 2.06.057 identify evidence of loss of material due to corrosion.

These inspections will be performed before the end of the current operating term and at intervals thereafter during the period of extended operation.

Results of the initial evaluations will be used to determine the appropriate inspection interval to ensure aging effects are identified prior to loss of intended function.12 Implement the Heat Exchanger Monitoring Program June 8, 2012 Letters B.1.15/as described in LRA Section B.1.15. 2.06.003 Audit Item and 320 2.06.057 2 COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section No./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 1 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 B.1.21. and 320 2.06.057 18 Enhance the Oil Analysis Program to periodically June 8, 2012 Letters B.1.22/change CRD pump lubricating oil. A particle count 2.06.003 Audit Item and check for water will be performed on the drained and 320 oil to detect evidence of abnormal wear rates, 2.06.057 contamination by moisture, or excessive corrosion.

19 Enhance Oil Analysis Program procedures for June 8, 2012 Letters B.1.22/security diesel and reactor water cleanup pump oil 2.06.003 Audit Item changes to obtain oil samples from the drained oil. and 320 Procedures for lubricating oil analysis will be 2.06.057 enhanced to specify that a particle count and check for water are performed on oil samples from the fire water 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. This includes 2.06.003 Audit Items destructive or non-destructive examination of one (1) and 219, 320 socket welded connection using techniques proven 2.06.057 by past industry experience to be effective for the identification of cracking in small bore socket welds.Should an inspection opportunity not occur (e.g., socket weld failure or socket weld replacement), a susceptible small-bore socket weld will be examined either destructively or non-destructively prior to entering the period of extended operation.

3 COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section No.Comments 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 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 1_ 2.06.057 24 Enhance the Service Water Integrity Program June 8, 2012 Letters B.1.28/procedure to clarify that heat transfer test results are 2.06.003 Audit Item trended. and 320 2.06.057 25 Enhance the Structures Monitoring Program June 8, 2012 Letters B.1.29.2 /procedure to clarify that the discharge structure, 2.06.003 Audit Items security diesel generator building, trenches, valve and 238, 320 pits, manholes, duct banks, underground fuel oil tank 2.06.057 foundations, manway seals and gaskets, hatch seals and gaskets, underwater concrete in the intake structure, and crane rails and girders are included in the program. In addition, the Structures Monitoring Program will be revised to require opportunistic inspections of inaccessible concrete areas when they become accessible.

26 Enhance Structures Monitoring Program guidance for June 8, 2012 Letters B.1.29.2 /performing structural examinations of elastomers 2.06.003 Audit Item (seals, gaskets, seismic joint filler, and roof and 320 elastomers) to identify cracking and change in 2.06.057 material properties.

27 Enhance the Water Control Structures Monitoring June 8, 2012 Letters B.1.29.3 /Program scope to include the east breakwater, jetties, 2.06.003 Audit Item and onshore revetments in addition to the main and 320 breakwater.

2.06.057 4 COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section No./.Comments 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).

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.

5

1. COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section No./Comments 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 following:

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 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.6

1. COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section NoJ Comments 34 Within the first 6 years of the period of extended June 8, 2018 Letter 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.

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-1 39 call on and General Electric SIL 644 Rev. 1. September 25, 2006 7 COMMITMENT IMPLEMENTATION SOURCE Related SCHEDULE LRA Section No./Comments 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 Xl 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.

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 License Renewal Application 12 (Letter 2.07.005), prior to the period of extended operation.

8 ATTACHMENT B to Letter 2.07.005 (3 pages)Response to Requests for Information Conveyed in NRC Letter Dated December 29, 2006 RAI Regarding Response to RAI 2.3.3.9-8:

The staff considers the response to Pilgrim RAI 2.3.3.9-8 incomplete.

Clarification of RAI 2.3.3.9-8 Response: During the conference call on December 12, 2006, it was recognized by the NRC license renewal staff that the fire suppression system for the three transformers adjacent to the Turbine Building was addressed in the Branch Technical Position 9.5-1 Appendix A response (Boston Edison Company letter dated March 9, 1977) and related staff safety evaluation report (dated December 21, 1978).Upon further consideration, automatic water spray systems to the main transformer, auxiliary transformer, and shutdown transformer are conservatively included in scope and subject to aging management review. The LRA is revised to add the following line items to Table 3.3.2-9: Piping Pressure Carbon Air- indoor Loss of System VII.1-8 3.3.1-58 A boundary steel (int) material walkdown (A-77)Nozzle Pressure Carbon Air -Loss of System VII.I-9 3.3.1-58 A boundary steel outdoor material walkdown (A-78)(ext)Nozzle Pressure Carbon Air- indoor Loss of System VII.1-8 3.3.1-58 A boundary steel (int) material walkdown (A-77)Valve Pressure Carbon Air- Loss of System VII.1-9 3.3.1-58 A body boundary steel outdoor material walkdown (A-78)(ext)RAI Regarding Response to RAI 2.4.3-1: The NRC staff feels RAI 2.4.3-1 trash racks and traveling screens require aging management programs; request that Entergy justify the RAI 2.4.3-1 jib crane being considered out of scope of license renewal.Clarification of Response to RAI 2.4.3-1: The trash racks and traveling screens are located in the flow path of seawater entering the Intake Structure.

The jib crane is located on the Intake Structure, above the trash racks.The trash racks are removable nonsafety-related components, constructed of epoxy coated steel, installed into the openings provided for the racks in the non-Class 1 portion of the Intake Structure.

The traveling screens are removable nonsafety-related components, constructed of epoxy coated steel frames, fiberglass baskets, and stainless steel basket mesh, carrying chain, sprockets, shafts, and bushings.

The screens are installed in the openings provided for the screens in the non-Class 1 portion of the Intake Structure.

Failure of the trash racks and traveling screens will not prevent accomplishment of the safety functions of the salt service water (SSW) system for the following reasons. A concrete baffle in front of the traveling screens protrudes down to below the surface of the seawater, arresting the transport of floating debris.The sluice gate openings into the SSW system pumps bays are below the surface of the water and above the floor of the Intake Structure.

In addition, during a design basis accident the I required flow through the Intake Structure to support operation of the SSW pumps is less than two percent of the normal flow through the traveling screens. At the required SSW flow rate, the Intake Structure design features described above prevent the transport of debris to the suction of the SSW pumps. Therefore, the traveling screens have no license renewal intended function and are not within the scope and subject to aging management review.The jib crane, also referred to as the trash rack rake, was installed to enable the mechanical removal of significant waterborne material (e.g. seaweed) from the trash racks. The usefulness of the trash rack rake has been limited, mostly due to the maintenance effort that became necessary relatively soon after initial use. Since then, the use of the trash rack rake has been discontinued and the removal of waterborne materials from the trash racks has been accomplished by divers. The installation of the trash rack rake was reported in the annual 10 CFR 50.59 report for 1993 (Boston Edison Company letter dated June 30, 1994). This modification did not affect the Class 1 portion of the Intake Structure or Class 1 systems.Therefore, the jib crane is not within the outside the subject to aging management for license renewal.RAI on Commitment to the Boiling Water Reactor Vessel Internals Program (BWRVIP): The staff requires the applicant to make a commitment to boiling water reactor vessel and internals program, BWRVIP-94 and BWRVIP-76.

Response to RAI on Commitment to the BWRVIP: Pilgrim is committed to BWRVIP-76 and BWRVIP-94.

BWRVIP-94 is the overall requirement for BWRVIP, and Entergy procedure ENN-DC-135, BWRVIP Inspection Program, reflects this requirement.

As stated in correspondence between the BWRVIP chairman and the NRC, Entergy is committed to the BWRVIP which includes the requirements of BWRVIP-76 and BWRVIP-94.1 2 Additionally, BWRVIP-76 is already included in the PNPS vessel internals program described in LRA Appendix B.1.8, BWR Vessel Internals.

The BWR Vessel Internals Program at PNPS is consistent with the program described in NUREG-1 801,Section XI.M9, BWR Vessel Internals, which specifies inspections in accordance with the applicable and approved BWRVIP guidelines.

Specifically, the scope of the AMP invokes the guidelines of BWRVIP-76 for inspection and evaluation of the core shroud.'Letter 97-461 from Carl Terry (BWRVIP Chairman) to Brian Sheron (NRC), "B WRVIP Utility Commitments to the BWRVIP," dated May 30, 1997 2 Letter 97-870 from Carl Terry (BWRVIP Chairman) to Brian Sheron (NRC), "BWRVIP Utility Commitments to the BWRVIP," dated October 30, 1997 2 RAI on One-Time Inspection of CASS Components:

Discuss the validity of using one-time inspections to assess reduction of fracture toughness in cast austenitic stainless steel (CASS) components.

Response to RAI on One-Time Inspection of CASS Components:

The LRA is revised to replace the aging management program of "One-Time Inspection" with"Inservice Inspection" for "valve bodies < 4" NPS" and aging effect "reduction of fracture toughness".

The affected line item is now consistent with Table 3.1.1, Item 55.LRA Section B.1.23, Program Description, where small bore piping in the reactor coolant system is discussed, is revised by replacing the phrase "cracking and reduction of fracture toughness are not occurring or are so insignificant" with "cracking is not occurring or is insignificant..." RAI B.1.3-2 on Exception to NUREG-1800 Section XI.M6: Discuss RAI B.1.3-2 exception to NUREG-1800 AMP XI.M6.Response to RAI B.1.3-2 on Exception to NUREG-1 800 Section XI.M6: The exception was included in the response to RAI B.1.3-2 provided in Attachment C of LRA Amendment 7 (Letter 2.06.079).

As a result of discussions held during the conference call on December 12, 2006 with the NRC license renewal staff, Entergy stated and the NRC staff agreed that the exception to NUREG-1800 Section XI.M6 need not be deleted.3 ATTACHMENT C to Letter 2.07.005 (2 pages)Supplemental Response to the Request for Additional Information on Response to RAI 4.3.1.2-2 Contained in LRA Amendment 9

Request for Supplemental Information on Response to RAI 4.3.1.2-2:

The loose part assessment provided in your October 6, 2006 response to RAI 4.3.1.2-2 is incomplete because it did not address the impact on the structural integrity of surrounding and downstream piping and components due to broken pieces other than big ring-type pieces.Please supplement your response by also addressing broken thermal sleeve pieces which may escape into the flow and damage inner thermal sleeve and downstream piping and components.

Response to Request for Supplemental Information on Response to RAI 4.3.1.2-2:

The October 6, 2006 response to RAI 4.3.1.2-2, in Attachment D of LRA Amendment 9 (Letter 2.06.089), in item (5) stated: "The thermal sleeve is not pressure boundary so its failure would not compromise the pressure boundary.

Failure of the thermal sleeve would be detected as a change in differential pressure of the affected jet pumps. There would be some slight movement but the thermal sleeve would remain within the nozzle. The movement of the riser pipe is restricted by the shroud. In addition, the cracks are at the outer end of the outer thermal sleeve. A full circumferential failure would not allow inward movement because the inner end of the outer thermal sleeve is welded to the nozzle and this would restrain movement." The response is supplemented as follows.Failure of the thermal sleeve by cracking until loose parts were generated is considered highly unlikely.

There is no operating experience of cracking to this extent. Cracking of the thermal sleeve to the point of separation would relieve the stress and limit further stress corrosion cracking.

Nonetheless, the improbable generation of loose parts would not result in safety concerns.

The following analysis of loose parts in the reactor internals was provided in BWRVI P-06-A.First, large pieces (over three inches in diameter) would become lodged in the jet pump nozzle.Such pieces could interfere with jet pump flow. BWRVIP-06, Section 4.1.1 states: "If blockage of the jet pump assembly or the recirculation pump suction were to occur, recirculation and jet pump flows would be affected and detected by routine operator surveillance.

If the effect is detectable, operator action could be expected to bring the plant to a safe condition.

Other settling locations would not impact the core flows and therefore would not present a safety concern." Second, small pieces (less than three inches in diameter) could pass through the jet pump nozzle and into the lower plenum.BWRVIP-06-A addresses loose parts in the lower plenum as follows: "In the lower plenum, the vertical component of the flow velocity is less than 10 ft/sec and would be insufficient to cause large parts to be lifted by the flow. Consequently, large loose parts which have been generated from lower plenum components would settle to the bottom head region where they would present no safety concern from the standpoint of fuel channel blockage.

Furthermore, the radial component of the flow velocity ranges from 7 ft/sec at the periphery to less than 1 ft/sec at the center and would tend to move the parts inward toward the reactor vessel centerline.

This inward movement would be restricted by the "forest" of control rod guide tubes in the lower plenum.I Small parts generated from lower plenum components or which have entered the lower plenum from the downcomer could be lifted by the flow and carried to the fuel bundle inlet orifice and the clearance between control rod guide tubes (about 1 -6 in.) is large enough to allow a small part to pass between them. The probability of the part negotiating a path through the "forest" of control rod guide tubes and finding its way to a fuel bundle orifice is considered small.Nevertheless, the vertical velocities in the lower plenum are sufficiently large that the small part might be carried toward a fuel support inlet orifice. The fuel support orifices range in size from about 1.2 to 2.4 inches in diameter, depending on location and specific plant design. Partial flow blockage of a fuel support inlet orifice can lead to initiation of boiling transition or possibly channel instabilities.

Due to the higher lift velocities and smaller orifice sizes in peripheral fuel bundles, small parts are more likely to block fuel channels in peripheral bundles than in central ones. Channel instability is less of a concern in this region due to the lower power distribution.

Smaller parts or debris that are able to pass through the inlet orifice could be stopped in the fuel bundle at the lower tie plate or the fuel rod spacer which have smaller clearances than the orifice. Because a blockage of one of these openings is smaller than that required to initiate boiling transition, there is no safety consequence of such small blockages." BWRVIP-06, Section 4.1.2 states: "access to the CRD guide tube by metallic parts is effectively prevented by the integrity of the guide tube and the core flow patterns which exist in the fuel bundle and bypass regions. Any debris which enters a CRD guide tube is unlikely to have sufficient mechanical strength to interfere with the operation of the CRD." 2 ATTACHMENT D to Letter 2.07.005 (1 page)Addition Information on Environmentally-Assisted Fatigue Supplemental RAI on Environmentally-Assisted Fatigue: Clarify whether Entergy factored in the oxygen concentrations derived from implementation of normal water chemistry (NWC) in the Fen calculations for those operational periods when NWC was being implemented instead of hydrogen water chemistry.

Response to Supplement RAI on Environmentally-Assisted Fatigue: For the license renewal application, environmentally assisted fatigue factors (Fens) were estimated based on hydrogen water chemistry oxygen concentration.

Prior to the period of extended operation, Pilgrim will perform fatigue analyses and appropriate Fens will be used, accounting for operating times with both hydrogen water chemistry and normal water chemistry.

License renewal commitment 31 in Attachment A has been modified to include this action.I ATTACHMENT E to LETTER 2.07.005 (1 page)Response to Request for Commitment to Groundwater Sampling Request for a Commitment to Groundwater Sampling: The NRC license renewal staff requests that a commitment be made for groundwater sampling to determine pH and the impact on concrete.Response to Request for a Commitment to Groundwater Sampling: The Structures Monitoring Program, described in LRA Section B.1.29.2, is revised to include enhancement that an engineering evaluation will be conducted periodically (at least once every five years) of groundwater samples to assess the aggressiveness of groundwater to concrete.Samples will be monitored for pH, chlorides, and sulfates.

License renewal commitment 43 governs implementation of this program enhancement.

License renewal commitments 25 and 26 pertain to other commitments made to enhance different aspects of the Structures Monitoring Program, and those commitments are not affected by this response or commitment.

LRA Amendment 5 (Letter 2.06.064, dated July 19, 2006) included the addition of the following sentence to LRA Section A.2.1.32, which is a summary description of the Structures Monitoring Program."License renewal comments 25 and 26 specify enhancements to this program." This sentence is changed as follows to include license renewal commitment 43: "License renewal commitments 25, 26, and 43 specify enhancements to this program." I