Text

Response to Request for Additional Information for Licence Renewal Application
	ML11145A090
Person / Time
Site:	South Texas
Issue date:	05/12/2011
From:	Gerry Powell; South Texas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NOC-AE-11002672, STI: 32862207
	Download: ML11145A090 (35)
	v • d • e

Nuclear Operating Company South Texas Project Electric GeneratingStation P0. Box 289 Wadsworth, Texas 77483 _V_ _--

May 12, 2011 NOC-AE-1 1002672 10CFR54 STI: 32862207 File: G25 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2746 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Response to Request for Additional Information for the South Texas Proiect License Renewal Application

Reference:

1. STPNOC Letter dated October 25, 2010, from G. T. Powell to NRC Document Control Desk, "License Renewal Application", (NOC-AE-1 0002607) (ML103010257)

2. NRC letter dated April 14, 2011, "Requests for Additional Information for the Review of the South Texas Project, License Renewal Application - Fire Protection and Component Integrity", (ML110830978)

By Reference 1, STP Nuclear Operating Company (STPNOC) submitted the License Renewal Application (LRA) for South Texas Project (STP) Units 1 and 2. By Reference 2, the NRC staff requested additional information for the review of the STP LRA. STPNOC's response to the request for additional information is included in the Enclosure 1 to this letter. contains a revision to regulatory commitment #30 in Table A4-1 of the LRA. There are no other regulatory commitments in this letter.

Should you have any questions regarding this letter, please contact either Arden Aldridge, STP License Renewal Project Lead, at (361) 972-8243 or Ken Taplett, STP License Renewal Project Regulatory point-of-contact, at (361) 972-8416.

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

Executed on t4, i*. 2011 Daie G. T. Powell Vice President, Technical Support & Oversight KJT : STPNOC Response to Request for Additional Information : List of Revised Licensing Commitments

NOC-AE-1 1002672 Page 2 N* cc:

(paper copy without enclosures) (electronic copy without enclosures)

Regional Administrator, Region IV A. H. Gutterman, Esquire U. S. Nuclear Regulatory Commission Kathryn M. Sutton, Esquire 612 East Lamar Blvd, Suite 400 Morgan, Lewis & Bockius, LLP Arlington, Texas 76011-4125 Balwant K. Singal John Ragan Senior Project Manager Catherine Callaway U.S. Nuclear Regulatory Commission Jim von Suskil One White Flint North (MS 8B1) NRG South Texas LP 11555 Rockville Pike Rockville, MD 20852 Ed Alarcon Senior Resident Inspector Kevin Polio U. S. Nuclear Regulatory Commission Richard Pena P. 0. Box 289, Mail Code: MNI 16 City Public Service Wadsworth, TX 77483 C. M. Canady Peter Nemeth City of Austin Crain Caton & James, P.C.

Electric Utility Department 721 Barton Springs Road C. Mele Austin, TX 78704 City of Austin John W. Daily Richard A. Ratliff License Renewal Project Manager (Safety) Alice Rogers U.S. Nuclear Regulatory Commission Texas Department of State Health Services One White Flint North (MS 011-Fl) 11555 Rockville Pike Rockville, MD 20852 Balwant K. Singal Tam Tran John W. Daily License Renewal Project Manager Tam Tran (Environmental) U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission One White Flint North (MS 011 F01) 11555 Rockville Pike Rockville, MD 20852

Enclosure 1 NOC-AE-1 1002672 Page 1 STPNOC Response to Request for Additional Information RAI 2.3.3.17-1

Background:

For South Texas Project (STP), Units 1 and 2, the staff reviewed the license renewal application (LRA); drawings; updated final safety analysis report (UFSAR), Section 9.5.1, and Fire Protection Evaluation and Comparison to Branch Technical Position (BTP), Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," May 1, 1976. The staff also reviewed the following fire protection documents cited in the current licensing basis listed in the STP, Units 1 and 2, "Operating License Conditions 2.E: Safety Evaluation Report" NUREG-0781, dated April 1986, and its Supplements.

Issue:

The staff has identified that instances of fire protection systems and/or components as noted in this request for additional information (RAI) have been excluded from the scope of license renewal and/or an aging management review (AMR). These systems/components that were not included in the license renewal boundaries appear to have fire protection intended functions required for compliance with Title 10 of the Code of FederalRegulations (CFR) 50.48, "Fire Protection," as stated in 10 CFR 54.4. Therefore, in order to complete our review, the staff requires a response to the following request.

Request:

The following LRA boundary drawings show the following fire protection systems/components as out of scope (i.e., not colored in green):

LRA Drawinq Systems/Components Location LR-STP-FP- Fire water suppression systems associated with Fire Protection 7Q271 F00046 transformers BOP 101 and 102 Loop LR-STP-FP- Fire water suppression system in the Lighting Fire Protection 7Q271 F00046 Diesel Generator Building Loop LR-STP-FP- Fire water suppression systems associated with Fire Protection 7Q272F00046 transformers BOP 201 and 202 Loop LR-STP-FP- Several fire water suppression systems associated B7 and D8 7Q272F00046 with various buildings, (e.g., Bldg. 15, Bldg. 27, Bldg. 33, Bldg. 45, Bldg. 50, Bldg. 52, and Bldg.71)

The staff requests that the applicant verify whether the fire protection systems/components listed above are in the scope of license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21(a)(1). If they are excluded from the scope of license renewal and are not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.

STPNOC Response LRA Drawing LR-STP-FP-7Q271 F00046:

Note typo in RAI, the transformer numbers are "1D1" and "1D2" not "101" and "102".

Enclosure 1 NOC-AE-1 1002672 Page 2 BOP transformers 1 D1 and 1D2 are non-safety and do not perform any license renewal related function. Therefore, these transformers are not within the scope of license renewal. The transformers are not located within 50 feet of a safety-related building. Therefore, fire suppression for the BOP transformers is not within the scope of license renewal and is not subject to aging management.

LRA Drawing LR-STP-FP-7Q271 F00046:

The fire water suppression system in the fire protection loop lighting diesel generator building was incorrectly omitted from the scope of license renewal as well as the lighting diesel generator. The lighting diesel generator provides power to outdoor lighting to illuminate access routes that may require operator travel to various safe shutdown components. STPNOC will amend the application to include the lighting diesel generator, the lighting diesel generator fuel supply, the lighting diesel generator building and the fire water suppression system.

LRA Drawing LR-STP-FP-7Q272F00046:

Note typo in RAI, the transformer numbers are "2D1" and "2D2" not "201" and "202".

BOP transformers 2D1 and 2D2 are non-safety and do not perform any license renewal related function. Therefore, these transformers are not within the scope of license renewal. The transformers are not located within 50 feet of a safety-related building. Therefore, fire suppression for the BOP transformers is not within the scope of license renewal and is not subject to aging management.

LRA Drawing LR-STP-FP-7Q272F00046:

Buildings 27, 33, 45, 50, 52 and 71 are located outside the protected area and contain no equipment important to safety. A fire in any of these buildings will not affect equipment or components important to safety. Therefore, fire suppression components in these buildings are not within the scope of license renewal.

Building 15 has been removed from the site. LRA drawing LR-STP-FP-7Q272F00046 will be updated to reflect this change.

RAI 2.3.3.17-2

Background:

For South Texas Project (STP), Units 1 and 2, the staff reviewed the license renewal application (LRA); drawings; updated final safety analysis report (UFSAR), Section 9.5.1, and Fire Protection Evaluation and Comparison to Branch Technical Position (BTP), Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," May 1, 1976. The staff also reviewed the following fire protection documents cited in the current licensing basis listed in the STP, Units 1 and 2, "Operating License Conditions 2.E: Safety Evaluation Report" NUREG-0781, dated April 1986, and its Supplements.

Issue:

The staff has identified that instances of fire protection systems and/or components as noted in this request for additional information (RAI) have been excluded from the scope of license renewal and/or an aging management review (AMR). These systems/components that were not included in the license renewal boundaries appear to have fire protection intended functions required for compliance with Title 10 of the Code of FederalRegulations (CFR) 50.48, "Fire Protection," as stated in 10 CFR 54.4. Therefore, in order to complete our review, the staff requires a response to the following request.

Enclosure 1 NOC-AE-1 1002672 Page 3 Request:

Section 9.5.1.2.1, Fire Protection Water Supply System," of the UFSAR on page 9.5-5, states that the water supply to refill the fire water storage tanks is normally provided from the Fresh Water System, which takes suction from a settling basin. This section also states that in the event of a failure in this system, the tank is refilled directly from the site well water system. LRA Section 2.3.3.17 discusses requirements for the fire water supply system but does not mention site well water pumps and associated components.

The staff notes that LRA boundary drawing LR-STP-FP-7Q270F00006 shows the site well water system and its components as out of scope (i.e., not colored in green). The staff requests that the applicant verify whether the site well water pumps and associated components to fire water storage tanks are in the scope of license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21(a)(1). The staff requested that, if these components are excluded from the scope of license renewal and are not subject to an AMR, the applicant provide justification for the exclusion.

STPNOC Response The fire water tanks are sized such that no makeup is required to meet the requirements for fire event safe shutdown. Site well water pumps and associated components are provided only for augmentation of storage tank capacity.

As stated in the STP FHAR section 4.1 page 4.1-7:

"STP has two separate dedicated water supplies consisting of two tanks of with 300,000 gallons (useable) each for fire protection water. The tanks are interconnected so that the fire pumps can take suction from either or both tanks. A failure of one supply will not result in a failure of the other supply.

The fire water supply is calculated on the basis of the largest expected flow rate for a period of two hours for safety-related areas or other areas that present a fire exposure hazard to safety-related areas. The largest expected flow rate has been calculated as follows: largest design demand of any water suppression system plus 500 GPM for manual hose streams for a minimum duration of two hours equals 296,000 gallons minimum water supply required."

Additionally, from STP UFSAR Section 9.5.1.2.1:

"Both (fire water storage) tanks are used only for fire water storage, and no outlets are provided for connection to any other system."

While the site well water system provides makeup to the fire water tanks, STP does not credit the refilling of the tanks to meet the requirements of 10 CFR 50.48 for fire event safe shutdown.

Therefore, the site well water system providing makeup to the fire water tanks is not within the scope of license renewal under the provisions of 10 CFR 54.4(a)(3).

RAI 2.3.3.17-3

Background:

For South Texas Project (STP), Units 1 and 2, the staff reviewed the license renewal application (LRA); drawings; updated final safety analysis report (UFSAR), Section 9.5.1, and Fire Protection Evaluation and Comparison to Branch Technical Position (BTP), Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," May 1, 1976. The staff also reviewed the following fire protection documents cited in the current licensing basis listed in the STP, Units 1 and 2, "Operating License Conditions 2.E: Safety Evaluation Report" NUREG-0781, dated April 1986, and its Supplements.

Enclosure 1 NOC-AE-1 1002672 Page 4 Issue:

The staff has identified that instances of fire protection systems and/or components as noted in this request for additional information (RAI) have been excluded from the scope of license renewal and/or an aging management review (AMR). These systems/components that were not included in the license renewal boundaries appear to have fire protection intended functions required for compliance with Title 10 of the Code of FederalRegulations (CFR) 50.48, "Fire Protection," as stated in 10 CFR 54.4. Therefore, in order to complete our review, the staff requires a response to the following request.

Request:

Tables 2.3.3-17 and 3.3.2-17 of the LRA do not include the following fire protection components:

fire hose stations, fire hose connections, and hose racks

floor drains for fire water

" dikes and curbs for oil spill confinement

components in reactor coolant pump oil collection system The staff requests that the applicant verify whether the fire protection components listed above are in the scope of license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21(a)(1). If they are excluded from the scope of license renewal and are not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.

STPNOC Response

Fire hose stations, fire hose connections, and hose racks:

The fire hose stations, fire hose connections and hose racks are within the scope of license renewal and subject to an AMR. The component type, "valve", as identified in LRA Table 2.3.3-17, is used to represent fire hose stations and fire hose racks. Each individual fire hose station and fire hose rack is included with the isolation valve in its fire water supply. For fire hose stations and hose racks within the scope of license renewal and subject to an AMR, the representative firewater isolation valve has been highlighted in the color "green".

" Floor drains for firewater:

Floor drains used for the removal of firewater are evaluated as component type "piping" and are identified in LRA Table 2.3.3-23 and Table 2.3.3-24 as components within the scope of license renewal and subject to an AMR.

Dikes and curbs for oil spill confinement:

Dikes and curbs for oil spill confinement are provided for oil-filled transformers. The dikes for the engineered safety features (ESF) transformers are evaluated as component type "concrete element" and are identified in LRA Table 2.4-7 as within the scope of license renewal and subject to an AMR. These dikes and curbs prevent the spreading of a fire that could affect equipment or components important to safety. Each ESF transformer is located in a separate diked pit sized to contain 100% of the transformer oil.

The STP Fire Hazards Analysis Report, Fire Areas 39-41, documents that, in the diesel generator building, the diesel fuel oil storage tank rooms are capable of containing the contents of the tank in the event of a tank rupture. The walls and floors of the tank rooms are evaluated as component type "concrete element" and are identified in LRA Table 2.4-3 as components within the scope of license renewal and subject to an AMR.

Enclosure 1 NOC-AE-1 1002672 Page 5 Components in reactor coolant pump oil collection system:

Components in the reactor coolant pump oil collection system are within the scope of license renewal and subject to an AMR. Reactor coolant pump oil collection system component types "tank", "valve" and "piping" are identified in LRA Table 2.3.1-2 as components within the scope of license renewal and subject to an AMR. The reactor coolant pump oil collection system components are shown highlighted in the color "green" as within the scope of license renewal on license renewal boundary drawings LR-STP-RC-5R379F05042#1 and LR-STP-RC-5R379F05042#2 for Units 1 and 2. Reactor coolant pump oil collection system component types, "flame arrestor" and "splash guard", are within the scope of license renewal, subject to an AMR and will be added to LRA Table 2.3.1-2 and LRA Table 3.1.2-2.

Table 2.3.1-2 Reactor Coolant System Component Type Intended Function Class 1 Piping <= 4in Pressure Boundary Closure Bolting Pressure Boundary Flame Arrestor Pressure Boundary Flow Element Leakage Boundary (spatial)

Indicator Leakage Boundary (spatial)

Structural Integrity (attached)

Insulation Insulate (Mechanical)

Piping Leakage Boundary (spatial)

Pressure Boundary Structural Integrity (attached)

Pump Pressure Boundary Rupture Disc Leakage Boundary (spatial)

Splash Guard Direct Flow Tank Leakage Boundary (spatial)

Pressure Boundary Structural Integrity (attached)

Tubing Leakage Boundary (spatial)

Pressure Boundary Structural Integrity (attached)

Valve Leakage Boundary (spatial)

Pressure Boundary Structural Integrity (attached)

Enclosure 1 NOC-AE-1 1002672 Page 6 Table 3.1.2-2 Reactor Vessel, Internals,and Reactor Coolant System - Summary of Aging Management Evaluation - Reactor Coolant System (Continued)

Component IIntended Material Environment Aging Effect Aging Management Program NUREG- Table 1 Notes Type Function Requiring 1801 Vol. Item Management 2 Item Closure Bolting PB Carbon Steel Borated Water Cumulative Time-Limited Aging Analysis IV.C2-10 3.1.1.07 A Leakage (Ext) fatigue damage evaluated for the period of extended operation Closure Bolting PB Stainless Borated Water Cracking Bolting Integrity (B2.1.7) IV.C2-7 3.1.1.52 B Steel Leakage (Ext)

Closure Bolting PB Stainless Borated Water Loss of preload Bolting Integrity (B2.1.7)

IV.C2-8 3.1.1.52 B Steel Leakage (Ext)

Closure Bolting PB Stainless Borated Water Cumulative Time-Limited Aging Analysis IV.C2-10 3.1.1.07 A Steel Leakage (Ext) fatigue damage evaluated for the period of extended operation Flame Arrestor PB Carbon Steel Lubricating Oil (Int) Loss of material Lubricating Oil Analysis VII.G-26 3.3.1.15 D,3 (B2.1.23) and One-Time Inspection (B2.1.16)

Flame Arrestor PB Carbon Steel Plant Indoor Air (Ext) Loss of material External Surfaces Monitoring V.C-1 3.2.1.31 D,3 Program (B2.1.20)

Flow Element LIB, Stainless Borated Water None None IV.E-3 3.1.1.86 A Steel Leakage (Ext)

Flow Element LB5 Stainless Treated Borated Loss of material Water Chemistry (B2.1.2) and V.D1-30 3.2.1.49 E, 2 Steel Water (Int) One-Time Inspection (B2.1.16)

Flow Element LBI Stainless Treated Borated Cracking Water Chemistry (B2.1.2) and V.D1-31 3.2.1.48 E, 2 Steel Water (Int) One-Time Inspection (B2.1.16)

Indicator LB1 3, SIA Stainless Borated Water None None IV.E-3 3.1.1.86 A Steel Leakage (Ext)

Indicator LB, 3, SIA Stainless Treated Borated Loss of material Water Chemistry (B2.1.2) and V.D1-30 3.2.1.49 E, 2 Steel Water (Int) One-Time Inspection (B2.1.16)

Enclosure 1 NOC-AE-1 1002672 Page 7 Table 3.1.2-2 Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation - Reactor Stainless Reactor Coolant (Int) Cracking ASME Section XI Inservice Steel Cast Inspection, Subsections IWB, Austenitic IWC, and IWD for Class 1 components (B2.1.1) and Water Chemistry (B2.1.2)

Pump PB Stainless Reactor Coolant (Int) Loss of fracture ASME Section XI Inservice IV.C2-6 3.1.1.55 A Steel Cast toughness Inspection, Subsections IWB, Austenitic IWC, and IWD (B2.1.1)

Pump PB Stainless Reactor Coolant (Int) Loss of material Water Chemistry (B2.1.2) IV.C2-15 3.1.1.83 A Steel Cast Austenitic Pump PB Stainless Reactor Coolant (Int) Cumulative Time-Limited Aging Analysis IV.C2-25 3.1.1.08 A Steel Cast fatigue damage evaluated for the period of Austenitic extended operation Rupture Disc LBS Stainless Borated Water None None IV.E-3 3.1.1.86 A Steel Leakage (Ext)

Rupture Disc LBS Stainless Treated Borated Loss of material Water Chemistry (B2.1.2) and V.D1-30 3.2.1.49 E, 2 Steel Water (Int) One-Time Inspection (B2.1.16)

Rupture Disc LBS Stainless Treated Borated Cracking Water Chemistry (B2.1.2) and V.D1-31 3.2.1.48 E, 2 Steel Water (Int) One-Time Inspection (B2.1.16)

Splash Guard DF Carbon Steel Plant Indoor Air (Int) Loss of material Inspection of Internal Surfaces VII.G-23 3.3. D, 3 in Miscellaneous Piping and 1.7 Ducting Components (B2.1.22) 1 Splash Guard DF Carbon Steel Plant Indoor Air (Ext) Loss of material External Surfaces Monitoring V.C-1 3.2. D,3 Procqram (B2.1.20) 1.3 1

Enclosure 1 NOC-AE-1 1002672 Page 8 RAI 2.4-1

Background:

For South Texas Project (STP), Units 1 and 2, the staff reviewed the license renewal application (LRA); drawings; updated final safety analysis report (UFSAR), Section 9.5.1, and Fire Protection Evaluation and Comparison to Branch Technical Position (BTP), Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," May 1, 1976. The staff also reviewed the following fire protection documents cited in the current licensing basis listed in the STP, Units 1 and 2, "Operating License Conditions 2.E: Safety Evaluation Report" NUREG-0781, dated April 1986, and its Supplements.

Issue:

The staff has identified that instances of fire protection systems and/or components as noted in this request for additional information (RAI) have been excluded from the scope of license renewal and/or an aging management review (AMR). These systems/components that were not included in the license renewal boundaries appear to have fire protection intended functions required for compliance with Title 10 of the Code of FederalRegulations (CFR) 50.48, "Fire Protection," as stated in 10 CFR 54.4. Therefore, in order to complete our review, the staff requires a response to the following request.

Request:

Section 2.4 of the LRA does not include the following fire barrier components:

Table 2.4-1, fire barrier seals

Table 2.4-4, concrete elements, concrete wall (masonry walls)

Table 2.4-8, fire barrier doors

Table 2.4-9, fire barrier coatings The staff requests that the applicant verify whether the above fire barrier components are in the scope of license renewal in accordance with 10 CFR 54.4(a) and subject to an AMR in accordance with 10 CFR 54.21(a)(1). If they are excluded from the scope of license renewal and not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.

STPNOC Response

Table 2.4-1, reactor containment building fire barrier seals:

There are no fire barrier seals within the scope of license renewal and subject to an AMR in the reactor containment building. The STP Fire Hazards Analysis describes the fire protection evaluation for the reactor containment building in FHAR Section 3.8. The reactor containment building is made up of a single fire area administratively divided into fire zones.

The zone boundaries consist of compartment walls and the primary and secondary shield walls with the exception of the peripheral zones, which are bounded by the containment exterior wall. The containment penetrations and the personnel access hatch, which are of heavy steel construction, are not 3-hour fire rated. Fire area drawings 7C-14-9M-24500 through 7C-14-9M-24505 show that there are no rated barriers separating the fire zones within the reactor containment building. Therefore, there are no fire barrier seals credited as performing a fire barrier function in the reactor containment building.

Table 2.4-4, turbine generator building concrete elements and concrete block (masonry walls):

Concrete elements and concrete block (masonry walls) with an intended function of "fire barrier" are within the scope of license renewal and subject to an AMR in the turbine

Enclosure 1 NOC-AE-1 1002672 Page 9 generator building. The intended function "fire barrier" will be added to component type "concrete elements" in LRA Table 2.4-4 and Table 3.5.2-4. Component type "concrete block (masonry walls)" with an intended function of "fire barrier" will be added to LRA Table 2.4-4, Section 3.5.2.1.4, and Table 3.5.2-4.

Table 2.4-4 Turbine GeneratorBuilding k.,..Te**me Funcni*#f '.Intended '

Concrete Elements ;Fire Barrier iFlood Barrier Shelter, Protection

.. Bl.

crt (Masonr Walls.'Structural Support 6C-oncieBIe-Bic_-

kFsoniWalisj ire Barrie.r Shelter, Protection

. . ... .. Structural Support The following information will be added to Section 3.5.2.1.4 of the LRA:

3.5.2.1.4 Turbine Generator Building Materials The materials of construction for the turbine generator building component types are:

Carbon Steel

Concrete

Concrete Block (Masonry Walls)

Elastomer Environment The turbine generator building component types are exposed to the following environments:

Atmosphere/Weather (Structural)

Buried (Structural)

Encased in Concrete

Plant Indoor Air (Structural)

Aging Effects Requiring Management The following turbine generator building aging effects require management:

Concrete crackinq and spalling

Cracking

Cracking due to expansion

Cracking, loss of bond, and loss of material (spalling, scaling)

Cracks and distortion

Increase in porosity and permeability, cracking, loss of material (spalling, scaling)

Increased hardness, shrinkage and loss of strength

Enclosure 1 NOC-AE-1 1002672 Page 10

Loss of material

Loss of material (spalling, scaling) and cracking Aging Management Programs The following aging management programs manage the aging effects for the turbine generator building component types:

Fire Protection (B2.1.12)

Masonry Wall Program (B2.1.31)

Structures Monitoring Program (B2.1.32)

Enclosure 1 NOC-AE-1 1002672 Page 11 Table 3.5.2-4 Containme, Building Component Intended Type Function Concrete Block F1, SH Concrete Plant Indoor Air Cracking Fire Protection (B2 (Masonry SS Block (Structural) (Ext) and Masonry Wall Walls) (Masonry Program (B2.1.31)

Wall§1 Concrete FB, SH, Concrete Plant Indoor Air Cracking due to Structures Monitori ng II1.A3-2 3.5.1.27 A Elements SS (Structural) (Ext) expansion Program (B2.1.32)

Concrete FB, SH, Concrete Plant Indoor Air Cracking, loss of Structures Monitoring II1.A3-9 3.5.1.23 A Elements SS (Structural) (Ext) bond, and loss of Program (B2.1.32) material (spalling, scaling)

Concrete FB, SH, Concrete Plant Indoor Air Increase in Structures Monitoring III.A3-10 3.5.1.24 A Elements SS (Structural) (Ext) porosity and Program (B2.1.32) permeability, cracking, loss of material (spalling, scaling)

Concrete FB, FLB, Concrete Plant Indoor Air Concrete cracking Fire Protection (B2.1.12) VII.G-28 3.3.1.65 B Elements SH, SS (Structural) (Ext) and spalling and Structures Monitoring Program (B2.1.32)

Concrete FB, FLB, Concrete Plant Indoor Air Loss of material Fire Protection (B2.1.12) VII.G-29 3.3.1.67 B Elements SH, SS (Structural) (Ext) and Structures Monitoring Program (B2.1.32)

Notes for Table 3.5.2-5:

Standard Notes:

A Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP is consistent with NUREG-1 801 AMP.

B Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP takes some exceptions to NUREG-1 801 AMP.

C Component is different, but consistent with NUREG-1801 item for material, environment, and aging effect. AMP is consistent with NUREG-1801 AMP.

E Consistent with NUREG-1801 for material, environment, and aging effect, but a different aging management program is credited or NUREG-1801 identifies a olant-specific aging management program.

Plant Specific Notes:

1 NUREG-1 801 does not orovide a line in which concrete masonry is insoectedi ner

'I the Fire Protection nroernm (2.1.12'.

NUREG-1 801 does not orovide a line in which concrete masonrv is inspected ner the Fire Protection nrooram (132 1 12)-

I

Enclosure 1 NOC-AE-1 1002672 Page 12 Table 2.4-8, fuel handling building fire barrier doors:

Fire barrier doors are within the scope of license renewal and subject to an AMR in the fuel handling building. Component type "fire barrier doors" will be added to LRA Table 2.4-8 and Table 3.5.2-8.

Table 2.4 Fuel HandlingBuilding Component Type Intended Function Fire Barrier Doors Fire Barrier Shelter, Protection

Enclosure 1 NOC-AE-1 1002672 Page 13 Table 3.5.2-8 Containments, Structures, and Component Supports - Summary of Aging Management Evaluation - Fuel Handling Building Component Intended Material Environment Aging Effect Aging Management NUREG- Table I Item Notes Type Function Requiring Program 1801 Vol.

Manageme nt 2 Item Fire Barrier FB SH Carbon Steel Plant Indoor Air Loss of material Structures Monitorinq III.A3-12 3.5.1.25 A Doors (Structural) (Ext) Pro-gram (B2.1.32)

Fire Barrier FB, SH Carbon Steel Plant Indoor Air Loss of material Fire Protection (B2.1.12) VII.G-3 3.3.1.63 B Doors (Structural) (Ext)

Enclosure 1 NOC-AE-1 1002672 Page 14 Table 2.4-9, essential cooling water structures fire barrier coatings/wraps:

There are no fire barrier coatings/wraps within the scope of license renewal and subject to an AMR in the essential cooling water structures. The STP Fire Hazards Analysis describes the fire protection evaluation for the essential cooling water structures in FHAR Section 3.6.

There are no fire barrier coatings/wraps credited in FHAR Section 3.6 for performing a fire barrier function in the essential cooling water structures.

Section 4.3.2.1 Reactor Pressure Vessel. Nozzles. Head and Studs RAI 4.3.2.1-1

Background:

Section 4.3.2.1, page 4.3-11, states that the STP, Units I and 2, reactor pressure vessel (RPV) heads were replaced in the Fall of 2009 and the Spring of 2010, respectively.

Issue:

The staff notes that some replacement RPV heads have detected fabrication defects prior to installation in nuclear plants. Therefore, the staff is seeking information regarding the nondestructive examination data in the replacement RPV heads at STP units.

Request:

(1) Discuss any recordable indication(s) detected in the new RPV heads in both units.

(2) Discuss how the indications, if any, will be monitored to the end of 60 years.

(3) Discuss measures that have been taken and the design features in the replacement RPV head to minimize the degradation in control rod drive mechanism penetration nozzles.

STPNOC Response (1) No relevant indications were identified during the pre-service inspection of the Unit 1 and 2 Replacement Reactor Vessel Closure Heads (RRVCHs).

(2) Currently, STP uses ASME Section Xl, 2004 Edition, no addenda for the In-Service Inspection (ISI) program. In 2008, the NRC endorsement of Code Case 729-1 required augmentation of the STP ISI program. Code Case 729-1 includes alternative examination requirements for PWR RRVCHs with nozzles having pressure-retaining partial-penetration (j-groove) welds. The inspections described below are in the Third Interval 10 yr ISI Program Plan.

" The RRVCHs (with nozzles and partial penetration welds of pressurized water stress corrosion cracking (PWSCC) resistant materials) are monitored by performing visual examinations every third refueling outage.

The RRVCH nozzles and partial penetration welds of PWSCC resistant materials are monitored by performing volumetric and/or surface examinations once per interval.

Monitoring will continue to be managed in accordance with the ISI program.

(3) Measures that have been taken and the design features provided in the RRVH to minimize degradation in CRDM penetration nozzles include:

Application of thermally treated Alloy 690 material for RRVH penetrations.

Application of Alloy 52 weld filler metal for J-groove welds.

Enclosure 1 NOC-AE-1 1002672 Page 15 Application of automatic J-groove welding technology (including water-cooling to improve stress distribution through the CRDM adapter wall).

The penetrations in the RRVCHs are fabricated from Alloy 690 material and welded in the forging with Alloy 52 weld filler material. Both the Alloy 690 material and the Alloy 52 weld filler material are recognized in the industry for their excellent resistance to PWSCC.

RAI 4.3.2.1-2 Backgqround:

Section 4.3.2.1, page 4.3-11, second paragraph, states that the closure flanges, studs and nuts are designed for 107 tightening cycles. A cycle consists of the reactor head removal and re-installation. It also states that the number of reactor head removals and re-installations to support refueling operations in 60 years is estimated to be 80-cycles.

Issue:

In the stress analysis of the closure flanges, studs and nuts, transient cycles are an input parameter to determine the fatigue usage factor. It is not clear which cycles described above are used in the fatigue usage factor calculation.

Request:

Clarify how many cycles (80, 107, or 187) were used in calculating the fatigue usage factor for closure flanges, studs and nuts.

STPNOC Response The fatigue usage factor analysis uses 107 tightening cycles for closure flanges, studs and nuts as required by the original and replacement head design specifications.

RAI 4.3.2.1-3

Background:

Section 4.3.2.1, page 4.3-11, second paragraph, states that the maximum usage factor based on the designed number of transient cycles in the closure flanges is 0.089.

Issue:

The staff is not clear as to why the fatigue usage factor for the RPV head closure flange is an order of magnitude lower than that of the stud inserts and studs. The staff also seeks clarification on the terminology of the "maximum usage factor."

Request:

(1) Discuss why the flanges have such a low usage factor when compared to the stud inserts (0.8852) and studs (0.3372).

(2) Discuss whether the maximum usage factor is the same as the cumulative usage factor (CUF) as specified in the ASME Code,Section III.

STPNOC Response

1) The most significant difference is the effect that bolting has on the alternating stress intensities of the components. For the stud threads and inserts, the 107 bolting cycles result in alternating stress intensities of 146 ksi and 125 ksi respectively. For the vessel and head flange, the bolting transient results in an alternating stress intensity of less than 17 ksi.

Enclosure 1 NOC-AE-1 1002672 Page 16

2) "Maximum usage factor" is the same as "cumulative usage factor" as specified in the ASME Code, Section II1.

RAI 4.3.2.1-4

Background:

Section 4.3.2.1, page 4.3-12, first paragraph, states that the corrosion analyses, fatigue analyses, and the ASME Section XI 48-year flaw growth analysis for the bottom mounted instrumentation nozzles and lower head repairs are valid for the period of extended operation.

Therefore these time-limited aging analyses (TLAAs) are dispositioned in accordance with 10 CFR 54.21(c)(1 )(i).

Issue:

The applicant stated without detailed discussion that the corrosion analyses, fatigue analyses and flaw growth analysis for the bottom mounted nozzles and lower head repairs are valid for the period of extended operation. The applicant needs to provide a detailed explanation as to why these analyses are valid for the period of extended operation.

Request:

Explain in detail how these analyses are valid for the period of extended operation per 10 CFR 54.21(c)(1 )(i).

STPNOC Response Fatigue crack growth analysis - The fatigue crack growth analysis assumes the numbers of transients equivalent to 48 years of operation by using 120% (48/40) of the design numbers of transients in UFSAR Table 3.9-8. Since this fatigue crack growth analysis covers an additional 48 years of operation from the repair date of the condition in 2003, the fatigue crack growth analysis is valid through the period of extended operation and this TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(i).

Fatigue usage factor analysis - The fatigue usage factor analysis assumes transients equivalent to 50 years of operation. This extends the validity of the fatigue usage factor analysis of the repaired Bottom-Mounted Instrumentation (BMI) penetrations from the repair date of the condition in 2003 through the period of extended operation. Therefore this TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(i).

Corrosion analysis - The corrosion analysis calculates a corrosion rate of 0.00153 inches/year.

The rate is doubled to give the diametrical corrosion rate of 0.00306-inches/year, or 0.153 inches in the 50 years from the repair date which extends through the period of extended operation. A calculation of the effects of the increase in the head bore of the BMI nozzles due to base metal corrosion shows that the bore for the BMI half-nozzle repair/replacement can increase from 1.562 inches to 1.95 inches (a diametric increase of 0.388 inch] due to material loss and still meet the stress requirements of ASME Code Section II1. Therefore, this TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(i).

The application of the corrosion rate through the period of extended operation is conservative because general corrosion will significantly decrease after a period of time due to the lack of oxygen, tight geometry, and lack of RCS flow at the location.

The corrosion rate was calculated using the methodology documented in CE NPSD-1 198-P Rev. 0, which was reviewed by the NRC and a safety evaluation was issued. Site-specific

Enclosure 1 NOC-AE-1 1002672 Page 17 calculations were performed as required by the safety evaluation of the topical report.

Information concerning the effects of corrosion on the BMI half-nozzle repairs was previously provided to the staff in References 1 and 2 in support of relief request RR-ENG-2-33. This relief request was approved in Reference 3.

References:

1. STPNOC Letter NOC-AE-03001559. Steven E. Thomas, Manager, Plant Design Engineering; to the US NRC Document Control Desk. "South Texas Project Unit 1, Docket No. STN 50 498, Response to Request for Additional Information Regarding Request for Alternative RR-ENG-2-32 (TAC No. MB9696)." July 3, 2003. [ML031920109]

2. STPNOC Letter NOC-AE-03001563. Mark E. Kanavos, Manager, Design Engineering; to the US NRC Document Control Desk. "South Texas Project Unit 1, Docket No. STN 50-498, Supplemental Response to Request for Additional Information Regarding Request for Alternative RR-ENG-2-32 (TAC No. MB9696)." July 17, 2003. [ML032020109]

3. US NRC Letter. Robert A. Gramm, Chief, Section 1, Project Directorate IV, Division of Licensing Project Management, Office of Nuclear Reactor Regulation; to James J.

Sheppard, President and Chief Executive Officer, STP Nuclear Operating Company.

"South Texas Project, Unit 1 - Relief Request RR ENG 2-33, Alternative Flaw Characterization Criteria for Two Bottom-Mounted Instrument Penetration Welds (TAC No.

MB9727)." August 1, 2003. [ML032130454]

RAI 4.3.2.1-5 Back-ground:

Section 4.3.2.1, page 4.3-12, second paragraph, states that the fatigue analyses for the replacement reactor vessel closure head are valid for the period of extended operation.

Therefore, this TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(i).

Issue:

The applicant did not explain why the fatigue analyses for the replacement reactor vessel closure head are valid for the period of extended operation. Also, the staff is not clear whether the "fatigue analyses" imply "fatigue usage factor calculations" or "fatigue crack growth calculations."

Request:

(1) Clarify how the fatigue analyses for the replacement reactor vessel head are valid for the period of extended operation per 10 CFR 54.21(c)(1)(i).

(2) Confirm that the "fatigue analyses" referred to above are the fatigue usage factor calculations, not the fatigue crack growth calculations.

STPNOC Response (1) The STP Unit 1 and 2 Control Rod Drive Mechanism (CRDM) pressure housings, the Core Exit Thermocouple Nozzle Assemblies (CETNAs), and the Internal Disconnect Devices (IDDs) were replaced with the Replacement Reactor Vessel Closure Heads (RRVCHs) in 2009 and 2010, respectively. The new CRDMs and CETNAs were qualified for 40 years, which extends the design lives beyond the period of extended operation. The renewed operating licenses for STP Units 1 and 2 will expire in 2047 and 2048, respectively.

(2) The "fatigue analyses" referred to above are the fatigue usage factor calculations.

Enclosure 1 NOC-AE-1 1002672 Page 18 RAI 4.3.2.1-6

Background:

Section 4.3.2.1, page 4.3-13, Table 4.3-3, Footnote (2) states that "...The 40-year design basis number of events should be sufficient for 60 years of operation; therefore the calculated 40-year usage factors will not be exceeded..."

Issue:

The wording "should be" causes the aforementioned statement to be ambiguous.

Reauest:

Submit the analysis to demonstrate that the 40-year design basis number of events is sufficient for 60 years.

STPNOC Response The term "should be" only refers to potential that STP will exceed the 40-year design basis number of events. The projections are provided in LRA Table 4.3-2. They demonstrate that the 40-year design basis numbers of events are sufficient for 60 years. The Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1 of the LRA) ensures that these transients, with the exception of the 107 tightening cycles discussed in RAI 4.3.2.1-2, remain below the 40-year design basis number of events for 60 years. The number of tightening cycles is determined by the number of refueling outages, which is currently limited to 80 refueling outages.

RAI 4.3.2.1-7

Background:

Section 4.3.2.1, page 4.3-13, Table 4.3-3, Footnote (3) states that 60-year projections are performed by multiplying the design basis CUFs by 1.5, which is equal to 60 years/40 years, unless otherwise noted. The applicant applied a factor 1.5 to the usage factors calculated for 40 years to project the usage factors for 60 years. This method assumes that the fatigue usage factor is directly proportional to the transient cycles used in the analysis.

Issue:

The staff is especially concerned about the 60-year usage factors for the closure stud hole inserts and closure studs because their 60-year usage factors are either exceeding the allowable or substantial.

Request:

Demonstrate that this linear extrapolation of using a factor 1.5 is appropriate or at a minimum conservative.

STPNOC Response This approach extrapolates the number of transients assumed in the 40 year design basis to 60 years. This has been shown to be conservative through operating history as shown in LRA Table 4.3-2. The calculation of CUF is specified by ASME Section III paragraph NB-3222.4(e)(5). The assumed number of transients only affects Step 4 and subsequent steps of the ASME section paragraph. If the projected number of transients, n, is increased by 1.5 times, then the CUF, which is equal to the summation of the assumed numbers of transients divided by the numbers of transients allowed by the alternating stress intensity, N, will also be increased by 1.5 times as shown below:

CUF 40= Yni/Ni; CUF_60= 1(1.5 x ni) / Ni = 1.5 x CUF_40

Enclosure 1 NOC-AE-1 1002672 Page 19 If the calculated 60 year CUF approaches 1.0, the fatigue usage factor analysis will be managed through the Metal Fatigue of Reactor Coolant Pressure Boundary program and the TLAA dispositioned per 10 CFR 54.21(c)(1)(iii). If the calculated 60 year CUF demonstrates margin to Code allowable CUF of 1.0, then the TLAA is dispositioned per 10 CFR 54.21(c)(1)(ii).

This response also addresses RAI 4.3.2.4-5.

RAI 4.3.2.1-8

Background:

The plates and welds in the RPV and associated nozzles, flanges and supports may contain fabrication defects or service induced flaws.

Issue:

The staff seeks to understand the existing condition of the RPV materials in terms of flaws/indications remaining in service, if any, and to determine how the applicant will manage these flaws/indications during the period of extended operation.

Request:

Identify any flaws/indications that are remaining in service in the RPV, nozzles, flanges, or vessel supports. Discuss how these flaws/indications will be managed to the end of 60 years.

STPNOC Response To date, the only flaws remaining in service in the reactor pressure vessel are the flaws in Unit 1 BMI nozzles 1 and 46. The response to RAI 4.7.2-1 request 1 lists the sources searched to identify remaining flaws.

Section 4.3.2.4 Pressurizer and Pressurizer Nozzles RAI 4.3.2.4-1

Background:

Section 4.3.2.4, page 4.3-17, first paragraph, states that pressure-retaining and support components of the pressurizer are subject to an ASME Section III fatigue analysis.

Issue:

This analysis has been revised for plant modifications with redefined loads, and for newly-identified design basis events not included in the original analyses.

Request:

(1) Describe in detail how the plant modifications affect the loads on the pressurizer and associated components (closures, nozzles, heaters, and support skirts).

(2) Discuss the redefined loads and newly-identified design basis events not included in the original analyses.

(3) Page 4.3-18 discusses the new loads and design basis events but without details. For example, where the cold overpressure mitigation system transient comes from? Where the 6000 pressure fluctuations come from and what is their impact?

STPNOC Response (1) The effects of the pressurizer weld overlay plant modifications are discussed in LRA Section 4.3.2.4. Other plant modifications evaluated are Thot reduction, replacement

Enclosure 1 NOC-AE-1 1002672 Page 20 steam generators, and reactor thermal power uprate. These modifications did not affect the loads on the pressurizer and associated components.

(2) The newly-identified design basis events are the cold overpressure mitigation system (COMS) actuation and the insurge-outsurge events. They are discussed in LRA Section 4.3.2.4. The term "redefined loads" refers to loads resulting from either the plant modification discussed above or the new design basis events.

(3) Cold overpressure mitigation system is used to satisfy the Technical Specification LCO 3.4.9.3 requirement for low temperature overpressure protection (LTOP). It was not initially incorporated into the Code fatigue usage factor analyses. The 6000 pressure fluctuation segment of the new transient definition, as defined by the NSSS vendor, is incorporated into the Code design specification. The impact of the pressure fluctuations is discussed in the response to RAI 4.3.2.4-3.

RAI 4.3.2.4-2

Background:

Section 4.3.2.4, page 4.3-17, Table 4.3-4 shows that some cumulative fatigue factors exceed the ASME Code,Section III allowable of 1.0.

Issue:

The staff is concerned about how the applicant will manage the aging effects of the components that have projected fatigue usage factor greater than 1.0 during the period of extended operation.

Request:

Describe in detail how the metal fatigue of reactor coolant pressure boundary aging management program (AMP) will manage those pressurizer components that have been projected to have a CUF exceeding the ASME Code,Section III, allowable of 1.0 at the end of 60 years. For example, discuss how the operator would know when the fatigue usage factor of a particular pressurizer component approaches 1.0, and the actions that would be taken to assess the components.

STPNOC Response The metal fatigue of reactor coolant pressure boundary program monitors the number of actual plant transients to ensure they do not exceed the number of transients used in the design fatigue analyses for the pressurizer components. The pressurizer design basis was reviewed to ensure that it was within the scope of the AMP or the program was enhanced to consider the additional transients included in the pressurizer design basis. Therefore, the monitoring program also ensures that the number of transient cycles experienced by the plant will be within the cycles used in the pressurizer design basis during the period of extended operation.

The current procedure for the metal fatigue of reactor coolant pressure boundary AMP requires the control room to complete daily screening data sheets. If a transient occurs, a transient specific datasheet is completed to record the plant's conditions during the event. This process will be changed for the period of extended operation to run computer software to assess plant instrumentation data recorded by the plant process computer and identify the transients that have occurred. At least once per fuel cycle, the information will be validated to ensure an accurate transient count and the actual transient severity remains within the design basis. The cycle counts are then compared to the action limits, and corrective action is initiated when a transient exceeds 80% of its design limit.

Enclosure 1 NOC-AE-1 1002672 Page 21 Corrective actions are discussed in more detail in LRA B3.1 and in the response to RAI 4.3.2.11-3, request 2.

This response also addresses RAI 4.3.2.4-6, request 3.

RAI 4.3.2.4-3

Background:

Section 4.3.2.4, page 4.3-18, first paragraph, states that "...The stress reports evaluated the effect on the pressurizer of 10 cold over-pressurization mitigation system activation events. The contribution of these thermal effects to the fatigue usage can be neglected..."

Issue:

The staff is not clear regarding how the applicant performed the stress analyses to show that the 10 cold over-pressurization mitigation system activation events can be neglected in the fatigue usage factor calculation.

Request:

(1) Discuss the stress reports.

(2) Explain why the contribution of these thermal effects to the fatigue usage can be neglected.

STPNOC Response (1) The effects of 10 cold over-pressurization mitigation system (COMS) activation events on the pressurizers were evaluated in an addendum to the stress reports. Many transient loadings are much more severe than the RCS over-pressurization event and the addition of the less-severe COMS transients in the design basis of the pressurizer had a minimal effect on the component Code analysis. The response to Request (2) below identifies these effects.

(2) COMS actuation is a pressure transient. The thermally-induced stresses associated with the transient are very small and the number of cycles of thermal events is low (10);

therefore, the contribution of these thermal effects to the fatigue usage factor can be neglected. To account for the pressure transient, the original fatigue usage factor contributions from the original design basis pressure transients were scaled to account for the additional cycles (6000) associated with the COMS pressure fluctuations.

This response also supports the response to RAI 4.3.2.4-1, request (3).

RAI 4.3.2.4-4

Background:

Section 4.3.2.4, page 4.3-19, first paragraph, states that "...The fatigue crack growth analysis for pressurizer spray, relief, safety, and surge nozzle preemptive overlays is not a TLAA because the crack is not qualified for the life of the plant, but only the inspection interval...

Since the crack is not qualified for the life of the plant, but only the inspection interval, the fatigue crack growth analysis is not a TLAA in accordance with 10 CFR 54.3(a), Criterion 3..."

Issue:

The pressurizer spray, relief, safety, and surge nozzles use nickel-based Alloy 82/182 dissimilar metal weld which is susceptible to primary stress-corrosion cracking (PWSCC). The applicant has installed weld overlays on the Alloy 82/182 welds to mitigate the potential for PWSCC. As part of weld overlay installation, the applicant is required to perform crack growth calculations to

Enclosure I NOC-AE-1 1002672 Page 22 project the growth of a postulated flaw or an actual detected flaw in the Alloy 82/182 weld. The staff is not clear why the applicant stated that the crack growth analysis is not a TLAA.

Request:

(1) Confirm that every overlaid Alloy 821182 welds in pressurizer spray, relief, safety and surge nozzles will be inspected every 10 years.

a. If not, the staff is concerned with the above statement regarding the fatigue crack growth analysis not being a TLAA. If an overlaid Alloy 82/182 is not inspected every 10 years, the fatigue crack growth analysis should be analyzed for the remaining life of the plant to ensure that the flaw will not affect the structural integrity of the pipe/weld to the end of 60 years.

b. In addition, the transient cycles should be monitored to verify that the fatigue crack growth analysis bounds the actual transient cycles. If this is true, then please justify why this cycle verification is not part of a TLAA per 10 CFR 54.21(c)(1)(iii).

(2) The staff understands that plant owners perform fatigue crack growth calculations for overlaid Alloy 82/182 welds using 40 years of transient cycles (sometimes 60 years).

Discuss how many years of transient cycles were used in the fatigue crack growth calculation for the overlaid Alloy 82/182 welds for various pressurizer nozzles.

(3) Discuss how the fatigue crack growth analyses for overlaid Alloy 82/182 welds at pressurizer spray, relief, safety and surge nozzles will be monitored to the end of 60 years (i.e., the end of the period of extended operation).

STPNOC Response (1) The overlaid Alloy 82/182 welds in pressurizer spray, relief, safety, and surge nozzles will be inspected every 10 years using a qualified PDI Ultrasonic technique in accordance with the ISI program and MRP-1 39/ASME Code Case N-770-1 requirements. The locations were inspected in Spring 2010 for Unit 2 (2RE14) and Fall 2009 for Unit 1 (1RE15) and no flaws were identified. The third ISI interval, which is scheduled to end in 2020, will adopt ASME Code Case N-770-1, "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation ActivitiesSection XI, Division 1"

a. The weld overlay will be inspected every 10 years.

b. The fatigue crack growth analyses were not identified as a TLAA thus they did not require a disposition. However the fatigue crack growth analyses which support the weld overlay work were performed with the same number of transients as the design fatigue analyses. These transients will be monitored by the Metal Fatigue of Reactor Coolant Pressure Boundary program.

(2) 40 years of transient cycles were used in the fatigue crack growth calculation.

(3) The fatigue crack growth analyses which support the weld overlay work were performed with the same number of transients as the design fatigue analyses. These transients will be monitored by the Metal Fatigue of Reactor Coolant Pressure Boundary program (AMP B3.1 described in the LRA).

RAI 4.3.2.4-5 Backqround:

Section 4.3.2.4, page 4.3-19, third paragraph, states that "...As shown in Table 4.3-4, the fatigue analyses of the safety and relief nozzles, and the seismic support lugs demonstrate

Enclosure 1 NOC-AE-1 1002672 Page 23 worst-case 40-year usage factors less than 0.4. When multiplied by 1.5 (60/40) to account for the 60-year period of extended operation, these results do not exceed 0.6, providing a large margin to the code acceptance criterion of 1.0..."

Issue:

The applicant used a ratio of 40 years to 60 years to project the fatigue usage factor from 40 years to 60 years. The staff is not clear if this direct proportional projection is acceptable.

The staff also questions a potential discrepancy in the fatigue usage factor value.

Request:

(1) The applicant applied a factor 1.5 to the usage factors calculated for 40 years to project the usage factors for 60 years. This method assumes that the usage factor is directly proportional to the transient cycles used in the analysis. Demonstrate that this linear extrapolation of using the ratio of 40 years vs. 60 years (60/40 = 1.5) is appropriate or at minimum conservative to project the 60-year usage factors.

(2) Table 4.3-4 of the LRA lists safety and relief nozzles as having a fatigue usage factor of 0.044 for 40 years and 0.066 for 60 years. The aforementioned statement cited the usage factors of 0.4 and 0.6. Clarify the discrepancy on the usage factors between Table 4.3-4 and the above statement.

STPNOC Response (1) The response to RAI 4.3.2.1-7 provides the justification for use of the 1.5 factor for fatigue dispositions.

(2) The usage factors for the safety and relief nozzles are 0.044 for 40 years and 0.066 for 60 years. The statement of "usage factors less than 0.4" is used in reference to the pressurizer seismic support lugs CUF as well, with a 40-year CUF of 0.29. The cited usage factors of 0.4 and 0.6 were chosen because they provide a large amount of margin to the Code limit of 1.0.

RAI 4.3.2.4-6

Background:

Section 4.3.2.4, page 4.3-19, last paragraph, states that "...The fatigue analyses of the remaining subcomponents have been found acceptable for a limiting number of transient events. The Metal Fatigue of Reactor Coolant Pressure Boundary program described in Section 4.3.1 and B3.1 ensures that the numbers of transients actually experienced during the period of extended operation remain below the assumed number; or that appropriate corrective actions maintain the design and licensing basis by other acceptable means..."

Issue:

The staff is not clear on the terminology used in the aforementioned paragraph. In addition, the staff is not clear how the metal fatigue AMP will monitor the transient cycles to verify the fatigue usage factor calculations.

Request:

(1) Identify the "remaining" subcomponents.

(2) Clarify whether the "fatigue analyses" are the fatigue crack growth analyses or cumulative fatigue usage factor analyses.

(3) Describe exactly how the metal fatigue AMP will monitor the transient cycles and track the fatigue usage factor calculations of components in Table 4.3-4.

Describe in detail the "appropriate corrective actions" and "acceptable means" in the above quoted statement.

Enclosure 1 NOC-AE-1 1002672 Page 24 STPNOC Response (1) The statement "remaining subcomponents" refers to all components in LRA Table 4.3-4 except the safety and relief nozzles and the seismic support lugs. The safety and relief nozzles and the seismic support lugs were dispositioned per 10 CFR 54.21 (c)(1)(i).

(2) The statement "fatigue analyses" is in reference to the cumulative fatigue usage factor analyses.

(3) The response to RAI 4.3.2.4-2 addresses how the metal fatigue AMP will monitor the transient cycles and track the fatigue usage factor calculations of components in Table 4.3-4 (4) The response to RAI 4.3.2.11-3, request (2) addresses the terms "appropriate corrective actions" and "acceptable means".

Section 4.3.2.11 Fatigue Crack Growth Assessments and Fracture Mechanics Stability Analyses for Leak-Before-Break (LBB) Elimination of Dynamic Effects of Primary Loop Piping Failures RAI 4.3.2.11-1

Background:

Section 4.3.2.11, page 4.3-31, states that "... NRC approval of the use of leak-before-break in the reactor coolant system primary loop piping was granted with STP SER, NUREG-0781, Supplement No. 2..." Section 4.3.2.1, page 4.3-32, states that "...Westinghouse determined that the conclusions of the previous LBB analysis for the reactor coolant piping, pressurizer surge line, and accumulator lines remain valid after steam generator replacement..."

Issue:

Based on these two statements, it is not clear exactly how many piping systems have been approved for LBB.

Request:

List the piping systems that have been approved for LBB and that are in the scope of the LRA.

STPNOC Response LBB analyses were performed for the reactor coolant piping, pressurizer surge line piping, safety injection accumulator piping, and the residual heat removal suction piping.

NRC approval of the use of leak-before-break analysis was granted with NUREG-0781, Supplements No. 2, 3 and 4.

RAI 4.3.2.11-2

Background:

Section 4.3.2.11, page 4.3-32, first paragraph, states that "...The Westinghouse LBB analysis for the primary loop cites a study which determined the effects of thermal aging on piping

Enclosure 1 NOC-AE-1 1002672 Page 25 integrity for a material at thermal embrittlement saturation. Therefore, the fracture mechanics evaluation is dependent on material properties not plant life and therefore, is not a TLAA by 10 CFR 54.3(a), Criterion 3..."

Issue:

The staff is not clear on why the cast austenitic stainless steel material in the LBB primary loop piping is not a TLAA even though the cast austenitic stainless steel material is susceptible to time-dependent thermal embrittlement degradation. If saturated fracture toughness is used in the LBB analysis, the LBB analysis of the cast stainless steel material could still be considered under 10 CFR 54.21(c)(i) or (ii). In addition, based on the aforementioned statement, the staff is not clear whether the applicant had used saturated fracture toughness in the LBB analysis.

Request:

(1) Clarify whether the saturated fracture toughness value due to thermal embrittlement was used in the fracture mechanics evaluation for LBB such that the fracture mechanics evaluation is not a TLAA.

(2) If the saturated fracture toughness is used in the fracture mechanics calculation (i.e., the fracture mechanics evaluation is dependent on material properties not plant life), explain why the fracture mechanics evaluation is not a TLAA.

(3) Cite or submit the reference of the study in the Westinghouse LBB analysis in the above quoted statement.

(4) Justify why the reference of the study has demonstrated the effects of thermal aging on cast austenitic stainless steel and that the material properties (i.e., fracture toughness) is at saturated condition.

STPNOC Response (1) The saturated fracture toughness value was used in all LBB analyses.

(2) Although the fracture mechanics calculation considers aging of the material property, aging is not based on the plant life. Aging is based on the minimum material properties possible and the value used by the calculation will be the same whether the plant life is 40 years, 60 years, or 100 years. Therefore, fracture mechanics calculation is not a TLAA in accordance with 10 CFR 54.3(a) Criterion 3.

(3) Westinghouse Report WCAP-10456. "The Effects of Thermal Aging on the Structural Integrity of Cast Stainless Steel Piping for Westinghouse Nuclear Steam Supply Systems."

November 1983.

(4) WCAP-1 0456 provides equations to predict end-of-life toughness for thermal aging of cast austenitic stainless steel materials based on silicon, chromium, molybdenum, and ferrite contents. Testing found that the material properties reached saturated conditions after 30,000 hours0 days 0 hours 0 weeks 0 months during a 60,000 hour0 days 0 hours 0 weeks 0 months test. Enclosure C, Item 2 of Reference 1, discusses the selection of fracture toughness properties.

Reference:

1. STP Letter ST-HL-AE-1617. J. H. Goldberg, VP, Nuclear Engineering and Construction, HL&P; to Harold R. Denton, Director, Office of Nuclear Reactor Regulation, USNRC.

"Alternative Pipe Break Criteria for Pressurizer Surge Line." March 12, 1986.

Enclosure 1 NOC-AE-1 1002672 Page 26 RAI 4.3.2.11-3

Background:

Section 4.3.2.11 page 4.3-32, mid-page, states that "...The metal fatigue of the reactor coolant pressure boundary program described in Section 4.3.1 and B3.1 ensures that the numbers of transients remain below the number actually experienced during the period of extended operation remain below the assumed number; or that appropriate corrective actions maintain the design and licensing basis by other acceptable means..."

Issue:

It appears that the applicant intended to state that the metal fatigue AMP will ensure that the numbers of transients used in the analysis remain higher than the actual numbers experience in operation. However, the staff is not clear whether the fatigue crack growth calculation in the LBB application will be part of the analysis that the metal fatigue AMP will be monitoring.

Request:

(1) Discuss whether the metal fatigue AMP specifically identifies all LBB analyses that will be monitored.

(2) Clarify the "appropriate corrective actions" and "other acceptable means" in the above statement.

(3) Clarify the aforementioned statement.

STPNOC Response (1) The transients used in the LBB analyses are consistent with those transients presented in LRA Table 4.3-2 with the exception of the following two transients not listed in LRA Table 4.3-2.

The first transient, "Accumulator Actuation, Accident Operation" is a combination of the "Inadvertent RCS Depressurization" transient and "LOCA" transient. The "LOCA" transient is a faulted event and therefore not counted. The "Inadvertent RCS Depressurization" transient listed in Table 4.3-2 is monitored and counted.

The second transient, "Reduce Temperature Return to Power" is identified in the pressurizer surge line fatigue crack growth analysis but not included in the STP design bases. This transient was designed to improve capabilities of the plant during load follow operations. STP does not practice load follow operations; therefore, this transient is not applicable to STP.

(2) This response also supports the responses to RAI 4.3.2.4-2 and RAI 4.3.2.4-6, request (4).

The term "other acceptable means" refers to actions other than counting cycles, which are meant to address fatigue at STP. When other acceptable corrective action is required, a 10 CFR 50.59 review is performed in order to determine if the methods and results are in line with the current licensing basis of STP or if regulatory review is needed.

The term "appropriate corrective actions" is in reference to the corrective action described in LRA Appendix B Section B3.1 and LRA Table A4-1, Commitment 30.

LRA Appendix B Section B3.1 Corrective Actions (Element 7) will be revised to read as follows:

Enclosure 1 NOC-AE-1 1002672 Page 27 Procedures will be enhanced to include appropriate cOrro.t.ve a+ctions to -he"n.... a component approachoS . a Y... coui..nt or CU ,actionlimit if a cy-cle co-u,,n+t a-cton limit is rea*ched, acceptable corretFie, atonn includo:

1) Review of fatigue usage calculations:

a) To identify the com.ponent. and analyses a.fetoed by tho transient in que.tion.

b) To dete-rn-me- whether the trani*ent iRnqueSton co.ntribhmutes sigFnificantly to CIUE.

G)To9 ensure that the analytical bases of the leak beeforee breaak (LB98) fatigue crack propagation analysis and of the high energy line break (HEmlB) locations are d) To ensure that thennal,+ical base- of a fatigue cGrak groth and .stabilityanalyis in suppot of relief frm AS-MbSr-m o-.-fig.n XI flawu removal

2) Evaluatio-n Of remaining marginso"n CU

3) Redefinition of the spec;fimd nummberFOf cyclei (e.g., by reducing specified numrbers f cycles for nther troansfnients And using the margiRn to incre.9ase the- al*lGWed numnber of cycles for th tranSi6ent that is appro.Ga*ing is specified number Of cycles).

4) Redefmi*nait of the transient to remov e.onservatim i* nthe.. presure and temperature FaRg9s.

Procedures will be enhanced to include appropriate corrective actions to be invoked if a component approaches a cycle count or CUF action limit. If a cycle count action limit is reached, acceptable corrective actions include:

(1) Review of fatigue usage calculations:

a) To identify the components and analyses affected by the transient in question.

b) To determine whether the transient in question contributes sigqnificantly to CUF.

c) To ensure that the analytical bases of the leak-before-break (LBB) fatigue crack propagation analysis and of the high ener-gy line break (HELB) locations are maintained.

d) To ensure that the analytical bases of a fatigue crack growth and stability analysis in support of relief from ASME Section Xl flaw removal.

(2) Evaluation of remaining margins on CUF.

(3) Review of fatigque crack growth and stability analyses support the leak before break exemptions and relief from the ASME Section Xl flaw removal or inspection requirements to ensure that the analytical bases remain valid. Re-analysis of a fatigue crack growth analysis must be consistent with or reconciled to the originally submitted analysis and receive the same level of regulatory review as the original analysis.

(4) Redefinition of the specified number of cycles (e.g., by reducing specified numbers of cycles for other transients and using the margin to increase the allowed number of cycles for the transient that is approaching its specified number of cycles).

(5) Redefinition of the transient to remove conservatism in the pressure and temperature ranges.

Enclosure 1 NOC-AE-1 1002672 Page 28 LRA Table A4-1, Commitment 30 will be revised to read as follows; Enhance the Metal Fatigue of Reactor Coolant Pressure Boundary program procedures to:

0 include additional locations nocessary to ensure accurate calculations of fatigue, 0 include additional transients that contribute siqnificantly to fatigue usage, 0 includo additional transients necessary to ensure accurat calculations of fatigue,fatigup uaguem toringtsonat specified locations, and specify frequency th e n and process periodic f reviews ofth eresults oftho monitored cycle count and CUFdata at least onc oeper fuel cycle,

include additional cycle count and fatigue usage action limits, which will invoke appropriate corrective actions ifa component approaches a cycle count action 'omit or a fatigue usage action limit. The acceptance criteria associated with the NUREG/CR 6260 sample locations a for newer vintage Westinghouse plant will c aafor uunt envirnmental er ad aeffects on fatiigue, 8include appropriate corrective actions to be invoked if a component approaches a cycle count action limit or a fatigue usage action limit.

Enhance the Metal Fatigue of Reactor Coolant Pressure Boundary program procedures to:

finclude additional locations necessary to ensure accurate calculations of fatigue,

" include additional transients that contribute significantly to fatigue usage, finclude counting of the transients used in the fatigue crack growth analyses, which support the leak-before-break analyses and ASME Section Xl evaluations to ensure the analyses remain valid,

( include additional transients necessary to ensure accurate calculations of fatigue, fatigue usage monitoring at specified locations, and specify the freguency and process of periodic reviews of the results of the monitored cycle count and CUE data at least once per fuel cycle.

" include additional cycle count and fatigue usage action limits, which will invoke appropriate corrective actions if a component approaches a cycle count action limit or a fatigue usage action limit. The acceptance criteria associated with the NUREGICR-6260 sample locations for a newer vintage Westinghouse plant will account for environmental effects on fatigue, and

" include appropriate corrective actions to be invoked if a component approaches a cycle count action limit or a fatigue usage action limit. Acceptable corrective actions include fatigue reanalysis, repair, replacement, or augmented inspections. Re-analysis of a fatigue crack growth analysis must be consistent with or reconciled to the originally submitted analysis and receive the same level of regulatory review as the original analysis.

(3) With the changes addressed in requests (1) and (2), the Metal Fatigue of the Reactor Coolant Pressure Boundary program will manage the LBB fatigue crack growth analysis in accordance with 10 CER 54.21 (c)(1)(iii).

RAI 4.3.2.11-4

Background:

SRP Section 3.6.3 specifies that LBB piping does not experience active degradation.

Licensees have used nickel-based Alloy 82/182 weld material in the LBB primary coolant piping. Alloy 82/1 82 welds are susceptible to PWSCC.

Enclosure 1 NOC-AE-1 1002672 Page 29 Issue:

The staff seeks to have a complete understanding of the Alloy 82/182 welds in LBB piping in the STP units and how the applicant manages PWSCC in the LBB piping during the period of extended operation.

Request:

(1) Identify the LBB pipes that are constructed using Alloy 82/182 filler weld metal which is susceptible to PWSCC.

(2) Identify the LBB pipes with Alloy 82/182 welds mitigated with weld overlays or mechanical stress improvement process.

(3) Identify the LBB pipes with unmitigated Alloy 82/182 welds. For this group of pipes, discuss the measures that will be implemented to minimize the potential for PWSCC.

(4) Was the LBB evaluation updated to account for any Alloy 82/182 welds in the piping system that were mitigated with weld overlay?

STPNOC Response (1) Units 1 and 2 reactor coolant piping and the pressurizer surge line are the LBB lines that contain Alloy 82/182 filler weld metal. Alloy 82/182 filler weld metal is used in the reactor vessel inlet and outlet nozzles, and the pressurizer surge nozzle.

(2) Mitigating structural weld overlays (SWOLs) were performed on the Alloy 82/182 filler weld metal in the Units 1 and 2 pressurizer surge lines. These mitigating SWOLs were performed on the pressurizer surge nozzles in Spring 2007 for Unit 2 and Fall 2006 for Unit

1. The locations were inspected in Fall 2009 for Unit 1 (1 RE15) and Spring 2010 for Unit 2 (2RE14) and no flaws were identified. These locations will continue to be inspected every 10 years using a qualified PDI Ultrasonic technique.

(3) Units 1 and 2 reactor coolant piping contains unmitigated Alloy 82/182 filler weld metal.

The inlet and outlet nozzles will be inspected with a qualified PDI Ultrasonic technique and in accordance with the ISI program and MRP-139/ASME Code Case N-770-1 requirements.

Ultrasonic testing was last performed during 1RE15/2RE14 and no flaws have been identified. The hot leg dissimilar metal welds are also visually inspected from the outside diameter every outage per Code Case N-722.

(4) The LBB evaluations for the STP Units 1 and 2 pressurizer surge lines were updated to account for the effects of primary water stress corrosion cracking (PWSCC) in the leak rate calculations. The results of the LBB evaluation show that the LBB margins (margin on leak rate of 10, margin of 2 on flaw size and a margin of 1 on loads by using absolute summation method of faulted loads combination) are satisfied. Therefore, the original analysis conclusions remain valid and the pressurizer surge line pipe breaks should not be considered in the structural basis of STP Units 1 and 2 after weld overlay application. This is discussed in LRA Section 4.3.2.10, "Subsection Increased CUF for Break Consideration."

In response to RIS 2010-007 "Regulatory Requirements for Application of Weld Overlays and Other Mitigation Techniques in Piping Systems Approved for Leak-Before-Break", an enhanced 10 CFR 50.59 evaluation is being performed. The enhanced 10 CFR 50.59 review will conclude that the methodology used for the STP updated LBB analysis is the same method the staff approved for use at Waterford Unit 3. A safety evaluation dated February 28, 2011, indicates that the analyses which consider the life of the plant are fatigue and crack growth analyses.

Enclosure 1 NOC-AE-1 1002672 Page 30 RAI 4.3.2.11-5

Background:

Section 4.3.2.4, pages 4.3-17 and 4.3-18, discuss new (redefined) loads and new design basis events due to plant modifications.

Issue:

The applicant discussed plant modifications which may result new pipe loads. The staff is not clear whether the applicant has considered the new loads in the LBB evaluations.

Request:

(1) Discuss whether the fracture mechanics calculations and fatigue crack growth calculations for all LBB piping have been updated to include these new loads and design basis events. If not, justify the validity of all LBB analyses in light of new loads and design basis events.

STPNOC Response LRA Section 4.3.2.11 includes a subsection "Effects of Power Uprate and Steam Generator Replacement on the LBB Analysis." The modifications reconciled the LBB analyses with the current plant design basis, including new loads. The evaluation determined that the conclusions of the previous LBB analyses for the reactor coolant piping, pressurizer surge line, and accumulator lines remain valid.

Section 4.7.2 In-Service Flaw Growth Analyses that Demonstrate Structural Stability for 40 Years RAI 4.7.2-1

Background:

The ASME Code,Section XI, IWA-3000 and IWB-3000, permit flaws to remain in service.

Small flaws may remain in service without performing analysis or successive examinations.

Larger flaws are required to be analyzed and examined in successive inspection periods.

Issue:

The applicant stated that it has performed a search and did not find any flaws evaluated for the remaining life of the plant other than those discussed elsewhere in the LRA. The staff seeks to understand how the applicant's search was performed. The staff also seeks to understand the flaws that are remaining in service but do not require evaluation. The purpose is to determine the overall condition of the pipes that are covered under the LRA.

Request:

Section 4.7.2, page 4.7-3, states that a search of the CLB [current licensing basis] did not identify any flaws evaluated for the remaining life of the plant other than those discussed elsewhere in this application. (1) Discuss the sources that have been searched to obtain this information. (2) Discuss whether there are recordable indications/flaws that have remained in-service in the piping without a flaw evaluation for pipes within the scope of license renewal application. Discuss how these flaws will be monitored to the end of 60 years.

STPNOC Response (1) The sources are the CLB document defined in LRA Section 4.1. They include:

" The Updated Final Safety Analysis Report (UFSAR)

" Technical Specifications

Enclosure 1 NOC-AE-1 1002672 Page 31

The NRC Safety Evaluation Reports (SERs) for the original operating licenses

Subsequent NRC Safety Evaluations (SEs)

STPNOC and NRC docketed licensing correspondence.

(2) There are no flaws remaining in service in the piping within the scope of license renewal application. A flaw was identified in the Unit 1 Refueling Water Storage Tank (RWST).

This is discussed below.

Unit 1 RWST had an indication of a small active leak at the top of the shell to base plate weld. Relief Request RR-ENG-33 was granted to operate with the flaw in place for one fuel cycle until the tank could be inspected [Reference 1]. The inspection indicated no evidence of baseplate or sidewall cracking inside the tank. Based on those inspection results and a large allowable flaw length of 63.6 inches, the NRC staff concluded that Unit 1 can continue to be operated, subject only to future inspections as required by Section Xl of the Code [Reference 2], which will monitor the leak to the end of 60 years.

Analysis of fatigue crack growth determined it was insignificant (growth of 1" for 100,000 fill/drain cycles) [Reference 3]. The safety evaluation of RR-ENG-33 [References 1 and 2]

found that the fatigue crack growth analysis is not required to be considered in the final safety determination, and thus is not a TLAA in accordance with 10 CFR 54.3(a) Criterion 4.

References:

1. NRC Letter. Robert A. Gramm, Chief, Section 1, Project Directorate IV and Decommissioning, Division of Licensing Project Management, Office of Nuclear Reactor Regulation; to Mr. William T. Cottle, President and Chief Executive Officer, STPNOC.

"South Texas Project, Unit 1 - Request for Relief from ASME Code Requirements Regarding Repair of Refueling Water Storage Tank with Flaw Indication (Relief Request RR ENG 33) (TAC No. MA7243)" June 22, 2000. [ML003725735]

2. NRC Letter. Mohan C. Thadani, Senior Project Manager, Section 1, Project Directorate IV, Division of Licensing Project Management, Office of Nuclear Reactor Regulation; to Mr.

William T. Cottle, President and Chief Executive Officer, STPNOC. "South Texas Project, Unit 1 - Evaluation of the Information Regarding the Examination by a Submerged Camera Inside the Refueling Water Storage Tank (TAC No. MB1413)." December 14, 2001.

[ML013460299]

3. STP Letter NOC-AE-00000481. "South Texas Project, Unit 1, Docket No. STN 50-498; Addendum to Request for Relief from ASME Boiler and Pressure Vessel Code Section XI Requirements (Relief Request RR-ENG-33)." February 22, 2000. [ML003686976]

Enclosure 2 NOC-AE-1 1002672 List of Revised Licensing Commitments

Enclosure 2 NOC-AE-1 1002672 Page 1 Table A4-1 License Renewal Commitments Commitment LRA Section Implementation Schedule 30 Enhance the Metal Fatigue of Reactor Coolant Pressure Boundary program procedures B3.1 Prior to the period of to: extended operation

include additional locations necessary to ensure accurate calculations of fatigue,

include additional transients that contribute significantly to fatigue usage,

" include counting of the transients used in the fatigue crack growth analyses, which support the leak-before-break analyses and ASME Section XI evaluations to ensure the analyses remain valid,

include additional transients necessary to ensure accurate calculations of fatigue, fatigue usage monitoring at specified locations, and specify the frequency and process of periodic reviews of the results of the monitored cycle count and CUF data at least once per fuel cycle,

" include additional cycle count and fatigue usage action limits, which will invoke appropriate corrective actions if a component approaches a cycle count action limit or a fatigue usage action limit. The acceptance criteria associated with the NUREG/CR-6260 sample locations for a newer vintage Westinghouse plant will account for environmental effects on fatigue, and

" include appropriate corrective actions to be invoked if a component approaches a cycle count action limit or a fatigue usage action limit. Acceptable corrective actions include fatigue reanalysis, repair, replacement, or augmented inspections. Re-analysis of a fatigue crack growth analysis must be consistent with or reconciled to the originally submitted analysis and receive the same level of reaulatorv review as the oriqinal analysis

v t e Letter Sequence Response to RAI
TAC:ME4936, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance... Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Meeting, Meeting, Meeting Questions 8 February 2012 22 May 2012 14 April 2011 14 April 2011 14 April 2011 24 May 2011 6 July 2011 12 July 2011 28 July 2011 15 August 2011 15 August 2011 15 August 2011 18 August 2011 21 September 2011 22 September 2011 22 September 2011 4 October 2011 11 October 2011 11 October 2011 18 October 2011 3 November 2011 6 December 2011 5 December 2011 14 December 2011 6 December 2011 4 January 2012 16 February 2012 9 February 2012 30 January 2012 28 February 2012 21 March 2012 19 April 2012 12 April 2012 25 June 2012 14 June 2012 26 July 2012 12 September 2012 3 October 2012 19 November 2012 18 December 2012 6 March 2014 16 July 2014 13 April 2015 27 May 2015 28 May 2015 11 June 2015 5 October 2015 30 November 2015 30 November 2015 22 April 2016 14 June 2016 18 January 2017 ... further results Answers 25 October 2011 31 May 2012 31 May 2012 31 May 2012 6 September 2016 5 May 2011 12 May 2011 23 June 2011 9 August 2011 23 August 2011 18 August 2011 15 September 2011 15 September 2011 10 October 2011 21 November 2011 21 November 2011 4 November 2011 21 November 2011 17 November 2011 6 December 2011 15 December 2011 15 December 2011 5 January 2012 18 January 2012 6 February 2012 16 February 2012 27 February 2012 12 March 2012 5 March 2012 28 March 2012 28 March 2012 29 March 2012 17 April 2012 26 April 2012 3 May 2012 31 May 2012 14 May 2012 31 May 2012 31 May 2012 7 June 2012 14 June 2012 5 July 2012 27 June 2012 17 July 2012 4 October 2012 21 August 2012 6 December 2012 27 February 2014 18 February 2014 20 March 2014 ... further results Results Approval, Approval Other: ML110620203, ML11140A163, ML11224A265, ML11230A003, ML11318A121, ML11318A122, ML11318A123, ML11318A124, ML11318A125, ML11318A126, ML11318A127, ML11318A128, ML11318A129, ML11318A130, ML11318A131, ML11318A132, ML11318A133, ML11318A135, ML11318A136, ML11318A138, ML11318A139, ML11318A140, ML11318A141, ML11318A142, ML11318A143, ML11318A144, ML11318A145, ML11318A146, ML11318A147, ML11318A148, ML11318A149, ML11318A150, ML11318A151, ML11318A152, ML11318A153, ML11318A154, ML11318A155, ML11318A156, ML11318A157, ML11318A158, ML11318A159, ML11318A160, ML11318A161, ML11318A162, ML11318A163, ML11318A164, ML11318A165, ML11318A166, ML11318A167, ML11318A168... further results
MONTHYEAR2011-06-16 [Table View]

NOC-AE-11002672, Response to Request for Additional Information for Licence Renewal Application

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NOC-AE-11002672, Response to Request for Additional Information for Licence Renewal Application

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