LR-N10-0313, Responses to Request for Additional Information Dated 08/03/10, Related to Bolting Integrity and 07/30/10 to Steam Generators and Question Posed During Region 1 Inspection, Regarding License Renewal Application

Responses to Request for Additional Information Dated 08/03/10, Related to Bolting Integrity and 07/30/10 to Steam Generators and Question Posed During Region 1 Inspection, Regarding License Renewal Application

ML102440676
Person / Time
Site:	Salem
Issue date:	08/26/2010
From:	Davison P Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N10-0313, TAC ME1834, TAC ME1836
Download: ML102440676 (21)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038 0 PSEG NuclearLLC AUG 26 2010 10 CFR 50 10 CFR 51 10 CFR 54 LR-N10-0313 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Nuclear Generating Station, Unit No. 1 and Unit No. 2 Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311

Subject:

Responses to 1) NRC Request for Additional Information, dated August 3, 2010, related to Bolting Integrity; 2) NRC Request for Additional Information, dated July 30, 2010, related to Steam Generators, and 3) A question posed during the August 2010 NRC Region I Inspection related to Steam Generators; all associated with the Salem Nuclear Generating Station, Units 1 and 2 License Renewal Application

References:

1. Letter from Ms. Bennett Brady (USNRC) to Mr. Thomas Joyce (PSEG Nuclear, LLC) "REQUEST FOR ADDITIONAL INFORMATION FOR SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION REGARDING BOLTING INTEGRITY (TAC NOS. ME1 834 AND ME1836)", dated August 3, 2010

2. Letter from Ms. Bennett Brady (USNRC) to Mr. Thomas Joyce (PSEG Nuclear, LLC) "REQUEST FOR ADDITIONAL INFORMATION FOR SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2 LICENSE RENEWAL APPLICATION (TAC NOS. ME 1834 AND ME 1836)," dated July 30, 2010

3. Discussion with Ms. Rachel Vaucher, USNRC, regarding Steam Generator Line Item In the reference 1 letter, the NRC requested additional information related to the submerged bolting associated with the Salem Nuclear Generating Station, Units 1 and 2 License Renewal Application. In the Reference 2 letter, the NRC requested additional information related to the Steam Generators. In the reference 3 discussions, an NRC/NRR reviewer posed a question related to a Steam Generator line item in the LRA, requesting clarification. Enclosed are the responses to these requests for additional information and request for clarification.

y/4/

Document Control Desk LR-N10-0313 Page 2 AUG 26 2010 This letter and its enclosure contain no new or revised regulatory commitments.

If you have any questions, please contact Mr. Ali Fakhar, PSEG Manager - License Renewal, at 856-339-1646.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on 0 d Ii Sincerely, Paul J. Davison Vice President, Operations Support PSEG Nuclear LLC

Enclosure:

Responses to NRC Requests for Additional Information and Request for Clarification cc: Regional Administrator - USNRC Region I B. Brady, Project Manager, License Renewal - USNRC R. Ennis, Project Manager - USNRC NRC Senior Resident Inspector - Salem P. Mulligan, Manager IV, NJBNE L. Marabella, Corporate Commitment Tracking Coordinator Howard Berrick, Salem Commitment Tracking Coordinator

Enclosure LR-N10-0313 Page 1 of 19 Enclosure Response to Requests for Additional Information related to Submerged Bolting and Steam Generators, and Request for Clarification associated with Steam Generators, related to the Salem Nuclear Generating Station, Units 1 and 2 License Renewal Application (LRA)

RAI B.2.1.9-02 RAI B.3.1.1.66-01 RAI B.3.1.1.84-01 Response to NRC Request for Clarfication Note: For clarity, portions of the original LRA text are repeated in this Enclosure.

Added text is shown in Bold Italics,and deletions are shown with strikethrough text.

Enclosure LR-N1 0-0313 Page 2 of 19 RAI B.2.1.9-02 Backaround:

The GALL Report AMP XI.M1 8, "detection of aging effects" states that for both ASME code class bolting and for other pressure retaining bolting, visual detection of leakage before the leakage becomes excessive is one acceptable way to inspect for loss of preload in bolting.

License renewal application Section B.2.1.9, Bolting Integrity, states that the program provides for managing cracking, loss of material, and loss of preload by performing visual inspections for pressure retaining bolted joint leakage in the following environments: air indoor and outdoor, raw water, and soil.

Issue:

It is not clear to the staff how visual inspection of bolting in a raw water environment would be capable of detecting loss or reduction of bolting preload since the environment would preclude capability to detect seepage or minor indications of leakage at bolted joints.

Requests:

a) Clarify what pressure joint bolting within the scope of the Bolting Integrity Program is exposed to an environment of raw water or treated borated water; and b) Explain how and on what frequency visual inspections are performed for submerged bolted joints and how those inspections would detect loss of preload in the bolting.

PSEG Response:

a) The population of pressure retaining bolted joints exposed to raw water within the scope of license renewal at Salem is limited to the service water pump bolting. The submerged portion of the Service Water pump casings includes bolted joints using stainless steel bolting material.

The only bolting at Salem that is within the scope of license renewal and is exposed to treated borated water is structural bolting in the Fuel Handling and Fuel Storage System. There are no pressure retaining bolted joints in the scope of license renewal exposed to a treated borated water environment at Salem.

During review of information in response to this RAI, it was identified that the carbon and low alloy steel bolting component type exposed to raw water in the Service Water system was inadvertently included within the scope of license renewal. The part of the Service Water system containing submerged carbon and low alloy steel bolting does not perform a license renewal intended function and is not within the scope of license renewal. Therefore, the associated bolting is also not within the scope of license renewal. As a result, LRA Table 3.3.2-23 is revised, as shown

Enclosure LR-N10-0313 Page 3 of 19 below, to delete the carbon and low alloy steel bolting exposed to a Raw Water external environment from the scope of license renewal.

Table 3.3.2-23 Service Water System Component Intended Material Environment Aging Effect Requiring Aging NUREG- Table 1 Notes Type Function Management Management 1801 Vol. Item Programs 2 Item 1Cr1tibon AGhQben-d -GhARRI Raw WateF LosA of Matorial/Genoral, Pitting, Belt4ng Glee-e Low .Ally Steel 4Em..a.) reVic.., and Micr.bieloGicall. IRntjrity Be!i nfl'-encod Corrocion Mecha.n'a.

.9Itig GaCrbonand Ran aw-A-ateU Loes of ProlkAd/'hermal Effects, Belti*

13egtue

-*- -

QG6eie Law Alloy Steel (Et-...a. ) Crep..d Self L ....

]Rket e§t44m'

_ _ _ _~~ Be"t~ _ ___ ___

b) Inspection of service water pump bolting will occur during the performance of the periodic service water pump inspection and repair procedure. The service water pumps are disassembled, inspected, repaired, and reassembled on a six year frequency. During disassembly, the pumps are inspected for loose or missing stainless steel bolting. The stainless steel service water pump bolting, exposed to a Raw Water external environment, is removed and inspected for loss of material during disassembly of the service water pumps. During reassembly of the service water pumps the bolting is torqued, as designed, to prevent loss of preload.

Performance of the service water pump inspection and repair procedure is sufficient to ensure loss of preload associated with service water pump bolting in a Raw Water external environment will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation.

Enclosure LR-N1O-0313 Page 4 of 19 For clarity, the LRA Appendix A, Section A.2.1.9, on page A-1 1 is revised as follows: (Note: LRA Appendix A, Section A.2.1.9 shown below includes the revision due to the response to RAI B.2.1.9-01 submitted in letter number LR-N10-0225, dated July 8, 2010)

A.2.1.9 Bolting Integrity The Bolting Integrity aging management program is an existing program that provides for aging management of pressure retaining bolted joints, component support bolting, and structural bolting within the scope of license renewal. The Bolting Integrity program incorporates NRC and industry recommendations delineated in NUREG-1339, "Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants," EPRI TR-104213, "Bolted Joint Maintenance & Applications Guide," and EPRI NP 5769, "Degradation and Failure of Bolting in Nuclear Power Plants," as part of the comprehensive corporate component bolting program. The program provides for managing cracking, loss of material and loss of preload of pressure retainiRg bolted joints by p....*,mneg vi.ual in.pection.for bolted joint leakage in the following environments: air, groundwater/soil, raw water, treated borated water, and soil.

Included in the aging management activities directedby this program are visual inspections for pressureretainingboltedjoint leakage and preventive measures.

The Bolting Integrity aging management program will be enhanced to include:

1. In the following cases, bolting material should not be reused:

a. Galvanized bolts and nuts,

b. ASTM A490 bolts; and

c. Any bolt and nut tightened by the turn of nut method.

This enhancement will be implemented prior to the period of extended operation.

Enclosure LR-N10-0313 Page 5 of 19 For clarity, the LRA Appendix B, Section B.2.1.9, Program Description on page B-51 is revised as follows: (Note: LRA Appendix B, Section B.2.1.9 shown below includes the revision due to the response to RAI B.2.1.9-01 submitted in letter number LR-N10-0225, dated July 8, 2010)

B.2.1.9 Bolting Integrity Program Description The Bolting Integrity aging management program is an existing program that provides for aging management of pressure retaining bolted joints, component support bolting, and structural bolting within the scope of license renewal. The Bolting Integrity program incorporates NRC and industry recommendations delineated in NUREG-1339, "Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants," EPRI TR-104213, "Bolted Joint Maintenance & Applications Guide," and EPRI NP 5769, "Degradation and Failure of Bolting in Nuclear Power Plants," as part of the comprehensive corporate component bolting program. The program provides for managing cracking, loss of material and loss of preload of prssurc retaining bolted joints by prfoFrming visual ...s.e,.c*tn f*o bolted joint l*akage in the following environments: air, groundwater/soil, raw water, treated borated water, and soil.

Included in the aging management activities performed by this program are visual inspectionsfor pressureretainingbolted joint leakage and preventive measures.

The Salem ISI program plan tables provide the examination category and description as identified in ASME Section XI, Table IWB-2500-1 for Class 1 components, Table IWC-2500-1 for Class 2 components, and Table IWD-2500-1 for Class 3 components.

Examinations are currently performed in accordance with the ASME Section Xl, 1998 edition up through and including 2000 addenda for Salem 1 and 2 per the Salem ISI program plans. Examinations for the period of extended operation will be in accordance with the appropriate code edition and addenda for the Salem ISI Program Plan. In accordance with 10 CFR 50.55a(g)(4)(ii), the program is updated each successive 120-month inspection interval to comply with the requirements of the latest edition of the ASME Code specified twelve months before the start of the inspection interval. The extent and schedule of the inspections is in accordance with IWB-2500-1, IWC-2500-1 and IWD-2500-1 and assures that detection of leakage or fastener degradation will occur prior to loss of system or component intended functions. Bolting associated with Class 1 vessel, valve and pump flanged joints receive VT-1 inspection. For other accessible pressure retaining bolting, routine observations will document any leakage before the leakage becomes excessive. Normally inaccessible system and component bolted joints will be inspected for degradation when made accessible during maintenanceactivities.

The integrity of accessible non-ASME Class 1, 2 and 3 system and component bolted joints, other than containment, is are evaluated by detection of visible leakage during maintenance or routine observation such as system

Enclosure LR-N10-0313 Page 6 of 19 walkdowns. Normally inaccessible non-ASME Class 1, 2, and 3 system and component boltedjoints will be inspected when made accessible during maintenanceactivities. Containment pressure retaining bolting is monitored using the ASME Section XI, Subsection IWE, B.2.1.28, aging management program.

High strength bolting material with actual yield strength > 150 ksi is used for NSSS Class 1 component supports. The bolts are installed in sliding connections, with no preload to allow for thermal movements. Aging management review determined that stress corrosion cracking (SCC) is not an applicable aging effect/mechanism since the bolts are not subject to high sustained tensile stress. The applicable aging effects are managed as part of the ASME Section Xl, Subsection IWF, B.2.1.30, aging management program.

Procurement controls and installation practices, defined in plant procedures, ensure that only approved lubricants, sealants, and proper torque are applied.

The activities are implemented through station procedures.

Other aging management programs also manage inspection of bolting and supplement this bolting integrity program. The ASME Section Xl Inservice Inspection (ISI) Subsections IWB, IWC, and IWD, B.2.1.1, aging management program manages the inspection of safety-related bolting and supplements this bolting integrity program. The ASME Section XI, Subsection IWF, B.2.1.30, aging management program addresses aging management of ASME Section Class 1, 2 & 3 piping and component support bolting. The ASME Section XI, Subsection IWE, B.2.1.28, aging management program addresses aging management of containment pressure retaining bolting. Other structural bolting is managed as part of the Structures Monitoring Program, B.2.1.33. The aging management of crane and hoist bolting is addressed by the Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems, B.2.1.13, aging management program. Aging Management of heating and ventilation bolted joints is addressed by the External Surfaces Monitoring, B.2.1.24, aging management program. The monitoring methods are effective in detecting the applicable aging effects and the frequency of monitoring has been adequate to prevent significant degradation. Inspection activities for bolting in a buried environment are performed in conjunction with buried piping and component inspections as part of the Buried Piping Inspection, B.2.1.22, aging management program or the plant specific Buried Non-Steel Piping Inspection, B.2.2.4, aging management program. Inspection activities for submerged bolting are performed in conjunction with associatedcomponent maintenanceactivities.

Class 1, 2 and 3 bolted joint repair falls within the scope of the ASME Section Xl Repair and Replacement Program. Flanged joint welding repairs are implemented in accordance with IWA-4000. Pressure bolting replacements are implemented in accordance with IWA-7000. Other pressure retaining bolting maintenance evaluations and repairs follow the EPRI bolting guidelines for the evaluation and repair of the flanges and replacement bolts. The ASME Section Xl, Subsection IWF, B.2.1.30, program addresses replacement of NSSS

Enclosure LR-N10-0313 Page 7 of 19 component support bolting. Corrective actions are addressed in accordance with 10 CFR Part 50, Appendix B.

The Bolting Integrity program will be enhanced, as noted below, to provide reasonable assurance that the aging effects will be adequately managed during the period of extended operation.

Enclosure LR-N10-0313 Page 8 of 19 RAI 3.1.1.66-01

Background:

The applicant states that the aging effect of loss of material due to erosion for steel steam generator secondary manways (cover only) exposed to air with leaking secondary side water and/or steam addressed in license renewal application (LRA)

Table 3.1.1, item 3.1.1-66 is not applicable. The applicant explains that these components are not exposed to air with leaking secondary side water and/or steam because there has been no operating experience at Salem Nuclear Generating Station (Salem) with leaking manways or handholds.

Issue:

The staff notes that even if the applicant has not observed any operating experience of leaking manways or handholes, it does not mean that the Material, Environment, Aging Effect/Mechanism, and aging management program combination can be totally and definitively excluded, especially during the extended period of operation, for these components. Therefore, the staff considers that the applicant has not provided enough information for justifying why LRA item 3.1.1.66 is not applicable.

Request:

Please demonstrate how the aging effect of loss of material due to erosion for steel steam generator secondary manways (cover only) exposed to air with leaking secondary side water and/or steam will never occur during the extended period of operation, or revise accordingly your proposed LRA Table 3.1.1, item 3.1.1-66.

PSEG Response:

Salem concurs with the staff that the aging effect of loss of material due to erosion for steel steam generator secondary manways (cover only) exposed to air with leaking secondary side water and/or steam may occur during the extended period of operation. This aging effect and mechanism also applies to the component type, Steam Generators (Inspection Ports and Diaphragm, Handholes and Covers) for the handhole covers only since they are also constructed of steel and are potentially exposed to the environment of air with leaking secondary side water and/or steam.

As a result of changes described above, LRA Table 3.1.1, Item Number 3.1.1-66 on page 3.1-38, is revised as follows. Revisions are indicated with bolded italics for inserted text and strikethroughs for deleted text.

It should be noted that Item Number 3.1.1-66 (GALL section IV.D2) pertains to the Once-Through Steam Generators (OSGs) section of GALL, Volume 1, Rev. 1, and Draft Rev. 2. The Salem Units 1 and 2 Steam Generators are Recirculating Steam Generators (RSGs), GALL section IV.D1.

Enclosure LR-N 10-0313 Page 9 of 19 Table 3.1.1 Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Aging Aging Further Item Number Component Effect/Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-66 Steel steam Loss of material due Inservice No Not applicablo. The -stelGotam generator to erosion Inspection, -R..ate,.

, ocondar.. manway,, and secondary Subsections handhold noo*..... r aro not exposod to a,-ir manways and (IWB, IWC, With a*k-in..g water and/r*

.Oc.ndary.side handholds (cover and IWD) for steam. There has been noe operat" only) exposed to Class 2 ,xprien-c.e.at Salem with leaking air with leaking components manway or9handholds,. Consistent secondary-side with NUREG-1801. The ASME Section water and/or Xl Inservice Inspection, Subsections steam IWB, IWC, and IWD program,B.2. 1.1, will be used to manage loss of materialdue to erosion for the steel steam generatorsecondarymanway and handhole covers exposed to air with leaking secondary-side water and/orsteam.

Enclosure LR-N 10-0313 Page 10 of 19 Additionally, LRA Section 3.1.2.1.4 on page 3.1-6, and LRA Table 3.1.2-4 on pages 3.1-132 and 3.1-140 are revised as shown below to incorporate the new environment, and aging effect and mechanism of Loss of Material due to Erosion for the component type, Steam Generators (Inspection Ports and Diaphragm, Handholes and Covers). Note that the aging effect and mechanism is only applicable to the carbon steel handhole covers.

The aging effect and mechanism of Loss of Material due to Erosion is also applied to the component type, Steam Generators (Secondary Manways and Covers), where only the manway covers are applicable. Revisions are indicated with bolded italics for inserted text.

3.1.2.1.4 Steam Generators Environments 0 Air with Leaking Secondary-side Water and/or Steam

Enclosure LR-N10-0313 Page 11 of 19 Table 3.1.2-4 Steam Generators Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Notes Type Function Requiring Programs Vol. 2 Item Item Management Steam Generators Pressure Carbon or Low Air with leaking Loss of ASME Section XI IV.D2-5 3.1.1-66 A (InspectionPorts Boundary Alloy Steel with secondary-sidewater Material/Erosion Inservice Inspection, and Diaphragm, StainlessSteel and/orsteam Subsections IWB, Handholesand Cladding IWC, and IWD Covers)

Steam Generators Pressure Low Alloy Steel Air with leaking Loss of ASME Section XI IV.D2-5 3.1.1-66 A (Secondary Boundary secondary-sidewater Material/Erosion Inservice Inspection, Manways and and/or steam Subsections IWB, Covers) IWC, and IWD

Enclosure LR-N10-0313 Page 12 of 19 RAI 3.1.1.84-01

Background:

The applicant states that the aging effect of cracking due to stress corrosion cracking (SCC) for nickel alloy steam generator components such as secondary side nozzles (vent, drain, and instrumentation) exposed to secondary feedwater and/or steam addressed in Table 3.1.1, item 3.1.1-84 is not applicable. The applicant explains that this component, material, environment, and aging effect/mechanism does not apply since Salem does not have nickel alloy steam generator secondary side nozzles exposed to secondary feedwater and/or steam.

Issue:

The staff notes that the applicant's description of its steam generators design in LRA Sections 2.3.1.4 and B.2.1.10, as well as in Updated Final Safety Analysis Report, Rev.

24, does not provide sufficient details about the materials of steam generator secondary side nozzles to judge whether the aging effect of SCC is really not applicable for those components.

Moreover, the staff notes that in LRA Table 3.1.2-4 related to steam generators, the applicant includes nickel alloy spray nozzles exposed to treated water but it does not address the aging effect of SCC.

Request:

Please clarify whether Salem steam generators contain any nickel alloy steam generator components exposed to secondary water/and/or steam, or revise accordingly your proposed LRA Table 3.1.1, item 3.1.1-84. Please justify why cracking due to SCC is an aging effect that does not need to be addressed for the nickel alloy steam generator spray nozzles exposed to treated water.

PSEG Response:

Salem inadvertently omitted the aging effect and mechanism of Cracking due to Stress Corrosion Cracking for the nickel alloy spray nozzles, which are installed in the Salem Unit 1 steam generators. Specifically, the LRA component type spray nozzles are the j-nozzles constructed of nickel alloy and are connected to each of the Salem Unit 1 carbon steel steam generator feedwater rings. Further, in response to RAI 3.1.2.2.14-01, in PSEG letter LR-N1 0-0271, dated July 28, 2010, the stainless steel material for the Unit 2 steam generator feedwater ring was added to LRA Table 3.1.2-4. The Salem Unit 2 j-nozzles are constructed of stainless steel and are connected to each of the Salem Unit 2 stainless steel feedwater rings.

It should be noted that Item Number 3.1.1-84 (GALL section IV.D2) pertains to the Once-Through Steam Generators (OSGs) section of GALL, Volume 1, Rev. 1, and Draft Rev. 2. The Salem Units 1 and 2 Steam Generators are Recirculating Steam Generators (RSGs), GALL section IV.D1.

Enclosure LR-N10-0313 Page 13 of 19 Therefore, LRA Table 3.1.1, Item Number 3.1.1-84 on page 3.1-44 is revised as follows to incorporate the applicable aging effects and mechanisms for the Unit 1 steam generator j-nozzles, referred to as component type, Spray Nozzles. Revisions are indicated with bolded italics for inserted text and strikethroughs for deleted text.

Table 3.1.1 Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Aging Aging Further Item Number Component Effect/Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-84 Nickel alloy steam Cracking due to stress Water Chemistry No Not appliable. This c.mpn..nt, generator corrosion cracking and One-Time m.atoial, -- yir.nmont, and aging components such Inspection or off ectlmohanism, doos not apply to as, secondary side ASME Inservice ,o-A,,to, V\1.sel, In.ternals, and Reac nozzles (vent, drain, Inspection (IWB, Ceolant Systo÷Ms Salom d4 not ha...

and instrumentation) IWC, and IWD). nickel alloy steam gonor.ator o...d4..

exposed to sido, no,-,-,-,s

.xpo.ed to.. cndary secondary feed'1-ator/steam. Consistent with feedwater/ steam NUREG-1801. The Water Chemistry program, B.2.1.2, and One-Time Inspection program, B.2.1.20, will be used to manage cracking due to stress corrosion cracking for the Salem Unit 1 nickel alloy steam generatorcomponents (i.e., the spray nozzles) exposed to secondary feedwateri steam (i.e., treated water).

Enclosure LR-N1 0-0313 Page 14 of 19 Additionally, LRA Table 3.1.2-4 on page 3.1-131 is revised to incorporate the new aging effect and mechanism of Cracking due to Stress Corrosion Cracking for the Salem Unit 1 component type, Spray Nozzles, constructed of nickel alloy. Due to the addition of the Salem Unit 2 stainless steel spray nozzles, LRA Table 3.1.2-4 is revised to include the aging effects and mechanisms of Cracking due to Stress Corrosion Cracking, and Loss of Material due to Pitting and Crevice Corrosion. The revisions to LRA Table 3.1.2-4 are shown below. Revisions are indicated with bolded italics for inserted text.

Enclosure LR-N10-0313 Page 15 of 19 Table 3.1.2-4 Steam Generators Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table 1 Notes Type Function Requiring Programs Vol. 2 Item Item Management Spray Nozzles Direct Flow Nickel Alloy Treated Water (Internal) Loss of Material/Pitting One-Time Inspection H, 2 (Unit 1) and Crevice Corrosion Spray Nozzles Direct Flow Nickel Alloy Treated Water (Internal) Loss of Material/Pitting Water Chemistry H, 2 (Unit 1) and Crevice Corrosion Spray Nozzles Direct Flow Nickel Alloy Treated Water Cracking/Stress One-Time Inspection IV.D2-9 3.1.1-84 A (Unit 1) (Internal) CorrosionCracking

> 140F Spray Nozzles Direct Flow Nickel Alloy Treated Water Cracking/Stress Water Chemistry IV.D2-9 3.1.1-84 A (Unit 1) (Internal) CorrosionCracking

>140 F Spray Nozzles Direct Flow Stainless Treated Water Loss of Material/Pitting One-Time Inspection VIII.F-23 3.4.1-16 A (Unit 2) Steel (Internal) and Crevice Corrosion Spray Nozzles Direct Flow Stainless Treated Water Loss of Material/Pitting Water Chemistry VIII.F-23 3.4.1-16 A (Unit2) Steel (Internal) and Crevice Corrosion Spray Nozzles Direct Flow Stainless Treated Water Cracking/Stress One-Time Inspection VIII.F-24 3.4.1-14 A (Unit2) Steel (Internal) CorrosionCracking

> 140F Spray Nozzles Direct Flow Stainless Treated Water Cracking/Stress Water Chemistry VIII.F-24 3.4.1-14 A (Unit2) Steel (Internal) CorrosionCracking I__________ _ 1_____ 1 >140 F I IIII

Enclosure LR-N10-0313 Page 16 of 19 Response to NFIC Request for Clarification from Region I License Renewal Inspection - August 2010 During the August 2010 NRC Region I License Renewal inspection, the Staff noted a discrepancy between LRA Table 3.1.2-4 and the Steam Generator Tube Integrity aging management program basis document. Component Types Steam Generator (Moisture Separators - Vanes and Dryers) and Steam Generator (Secondary Flow Distribution Baffle - Unit 1 only) are considered secondary internal components that were originally contained in the program basis document, but later removed from the LRA since they were determined to not have an intended function. Upon further review it has been concluded that the intended function of Structural Support will be applied to these Steam Generator secondary internal components. For both Units, the vanes and dryers are constructed of carbon steel. Only the Unit 1 Steam Generators contain the secondary flow distribution baffle, which is constructed of stainless steel. Additionally, the Steam Generator loose part monitoring (i.e., trapping) system on Unit 2 only, which is comprised of stainless steel components, and considered a secondary internal component, was not included in the LRA for aging management. This component also has the intended function of Structural Support. The Steam Generator secondary internals component type consist of the carbon steel and stainless steel moisture separators (dryers and vanes), the stainless steel Unit 1 secondary flow distribution baffle, and the stainless steel Unit 2 loose part monitoring (i.e., trapping) system.

As a result, Salem revises LRA Table 2.3.1-4 on page 2.3-32 as shown below to add a new component type, Steam Generators (Secondary Internals) and its intended function of Structural Support. Revisions are indicated with bolded italics for inserted text.

Table 2.3.1-4 Steam Generators Components Subject to Aging Management Review Component Type Intended Function Steam Generators(Secondary StructuralSupport Internals)

I Additionally, Salem revises LRA Table 3.1.2-4 as shown below to provide aging management line items for the added component type Steam Generator (Secondary Internals). Revisions are indicated with bolded italics for inserted text.

Enclosure LR-N10-0313 Page 17 of 19 Table 3.1.2-4 Steam Generators Component Intended Material Environment Aging Effect Requiring Aging Management NUREG-1801 Table 1 Notes Type Function Management Programs Vol. 2 Item Item Steam Generators StructuralSupport CarbonSteel Treated Water Loss of Material/General, Steam GeneratorTube IV.D1-12 3.1.1-16 E, 3 (Secondary (External) Pittingand Crevice Integrity Internals) Corrosion Steam Generators StructuralSupport CarbonSteel Treated Water Loss of Material/General, Water Chemistry IV.D1-12 3.1.1-16 A (Secondary (External) Pittingand Crevice Internals) Corrosion Steam Generators StructuralSupport CarbonSteel Treated Water Wall Thinning/Flow Steam GeneratorTube IV.D1-26 3.1.1-32 E, 3 (Secondary (External) Accelerated Corrosion Integrity Internals)

Steam Generators StructuralSupport StainlessSteel Treated Water Loss of Material/Pittingand One-Time Inspection VIII.F-23 3.4.1-16 A (Secondary (External) Crevice Corrosion Internals)

Steam Generators StructuralSupport Stainless Steel Treated Water Loss of Material/Pittingand Water Chemistry VIII.F-23 3.4.1-16 A (Secondary (External) Crevice Corrosion Internals)

Steam Generators StructuralSupport StainlessSteel Treated Water Cracking/StressCorrosion Steam GeneratorTube IV.D1-14 3.1.1-74 A (Secondary (External) Cracking Integrity Internals) > 140F Steam Generators StructuralSupport StainlessSteel Treated Water Cracking/StressCorrosion Water Chemistry IV.D1-14 3.1.1-74 A (Secondary (External) Cracking Internals) >140 F Table 3.1.1 is revised as shown below to add the new component type, Steam Generators (Secondary Internals) for GALL Line Item 3.1.1-16. Revisions are indicated with bolded italics for inserted text.

Enclosure LR-N10-0313 Page 18 of 19 Table 3.1.1 Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System I Aging Management Further Evaluation Discussion Item Aging Effect/Mechanism Programs Ecommen D Number Component Recommended 3.1.1-16 Steel steam generator upper Loss of material due to Inservice Inspection Yes, detection of Consistent with NUREG-1801. The ASME and lower shell and transition general, pitting and (IWB, IWC, and IWD), aging effects is to Section Xl Inservice Inspection, Subsections IWB, cone exposed to secondary crevice corrosion and Water Chemistry be evaluated IWC, and IWD program, B.2.1.1, will be used to feedwater and steam and, for verify the effectiveness of the Water Chemistry Westinghouse Model program, B.2.1.2, to manage the loss of material 44 and 51 S/G, if due to general, pitting and crevice corrosion in the general and pitting steel steam generator components (secondary corrosion of the shell manways and covers, tubesheets, upper. head, is known to exist, upper shell, conical shell, lower shell), piping additional inspection components and connections, and main feedwater procedures are to be and main steam nozzles exposed to steam and developed. treated water in the Steam Generators.

Components in the Steam Generators system have been aligned to this item number based on material, environment and aging effect. The Steam Generator Tube Integrity program, B.2.1.10, will be substituted to verify the effectiveness of the Water Chemistry program, B.2.1.2, to manage the loss of material due to general, pitting and crevice corrosion in the steel steam generator feedwater inlet ring and supports, and secondaryinternalsexposed to secondary feedwater and steam (i.e., treated water) for this system.

See subsection 3.1.2.2.2.2.

L 1 -.1. .~.

Enclosure LR-N 10-0313 Page 19 of 19 LRA Subsection 3.1.2.2.2.2, third paragraph, is revised as shown below. Revisions are indicated with bolded italics for inserted text.

3.1.2.2.2 Loss of Material due to General, Pitting, and Crevice Corrosion

2. Loss of materialdue to general,pitting, and crevice corrosioncould occur in the steel PWR steam generatorupper and lower shell and transition cone exposed to secondary feedwater and steam. The existing program relies on control of chemistry to mitigate corrosionand In-service Inspection (ISI) to detect loss of material. The extent and schedule of the existing steam generatorinspections are designed to ensure that flaws cannot attain a depth sufficient to threaten the integrity of the welds. However, accordingto NRC Information Notice (IN) 90-04, the program may not be sufficient to detect pitting and crevice corrosion,if general and pitting corrosionof the shell is known to exist. The GALL Report recommends augmented inspection to manage this aging effect. Furthermore, the GALL Report clarifies that this issue is limited to Westinghouse Model 44 and 51 Steam Generatorswhere a high stress region exists at the shell to transition cone weld. Acceptance criteriaare describedin Branch Technical Position RLSB- 1.

Salem will implement a Steam Generator Tube Integrity program, B.2.1.10, for susceptible locations to verify the effectiveness of the Water Chemistry program, B:2.1.2, to manage the loss of material due to general, pitting and crevice corrosion in the steel steam generator feedwater ring and supports, and secondaryinternals exposed to treated water in the Steam Generators. The Steam Generator Tube Integrity and Water Chemistry programs are described in Appendix B.