ML12097A063

Supplement to the Response to Requests for Additional Information for the South Texas Project License Renewal Application, Aging Management Program, Set 12

ML12097A063
Person / Time
Site:	South Texas
Issue date:	03/28/2012
From:	Rencurrel D South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-12002816, TAC ME4936, TAC ME4937, STI: 33389321
Download: ML12097A063 (19)

Text

Nuclear Operating Company South Texas Prolect Electric Generatinlg Station P.O Box 289 Wadsworth, Texas 77483 March 28, 2012 NOC-AE-12002816 10 CFR 54 STI: 33389321 File: G25 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Supplement to the Response to Requests for Additional Information for the South Texas Project License Renewal Application Aging Management Program, Set 12 (TAC Nos. ME4936 and ME4937)

References:

1. STPNOC letter dated October 25, 2010, from G. T. Powell to NRC Document Control Desk, "License Renewal Application" (NOC-AE-10002607) (ML103010257)

2. NRC letter dated February 8, 2012, "Requests for Additional Information for the Review of the South Texas Project, Units 1 and 2 License Renewal Application -

Aging Management, Set 12 (TAC Nos. ME4936 and ME 4937)"(ML12009A117)

3. STPNOC letter dated February 27, 2012, from D. W. Rencurrel to NRC Document Control Desk, "Response to Requests for Additional Information for the South Texas Project License Renewal Application Aging Management Program, Set 12 (TAC Nos. ME4937 and ME2937)" (NOC-AE-12002797) (ML12069A024)

4. STPNOC letter dated November 30, 2011, from D. W. Rencurrel to NRC Document Control Desk, "Annual Update to the South Texas Project License Renewal Application (TAC Nos. ME4936 and ME4937)" (NOC-AE-1 1002758)

(ML11335A140)

By Reference 1, STP Nuclear Operating Company (STPNOC) submitted a License Renewal Application (LRA) for South Texas Project (STP) Units 1 and 2. By Reference 2, the NRC staff requested additional information for review of the STP LRA. STPNOC provided a response to the requests for additional information in Reference 3. By Reference 4, STPNOC provided an annual update amendment to the LRA as required by 10 CFR 54.21 (b). This letter supplements the response provided in Reference 3 and revises annual update information in Reference 4. Enclosure 1 describes the supplemental response and revised annual update information. Changes to LRA pages described in Enclosure 1 are depicted as line-in/line-out pages provided in Enclosure 2.

There are no regulatory commitments provided 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.

AIL//I

NOC-AE-12002816 Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on bate Rencurrel Chief Nuclear Officer KJT

Enclosure:

1. STPNOC Supplemental Response to Requests for Additional Information

2. STPNOC LRA Changes with Line-in/Line-out Annotations

NOC-AE-1 2002816 Page 3 cc:

(paper copy)

(electronic copy)

Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, Texas 76011-4511 Balwant K. Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8B1) 11555 Rockville Pike Rockville, MD 20852 Senior Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 289, Mail Code: MN116 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 John W. Daily License Renewal Project Manager (Safety)

U.S. Nuclear Regulatory Commission One White Flint North (MS 011-Fl)

Washington, DC 20555-0001 Tam Tran License Renewal Project Manager (Environmental)

U. S. Nuclear Regulatory Commission One White Flint North (MS O11F01)

Washington, DC 20555-0001 A. H. Gutterman, Esquire Kathryn M. Sutton, Esquire Morgan, Lewis & Bockius, LLP John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Richard Pena City Public Service Peter Nemeth Crain Caton & James, P.C.

C. Mele City of Austin Richard A. Ratliff Alice Rogers Texas Department of State Health Services Balwant K. Singal John W. Daily Tam Tran U. S. Nuclear Regulatory Commission NOC-AE-12002816 Enclosure I STPNOC Supplemental Response to Requests for Additional Information NOC-AE-12002816 Page 1 of 5 Changes to correspondence reference 3:

Reference 3:

STPNOC letter dated February 27, 2012, from D. W. Rencurrel to NRC Document Control Desk, "Response to Requests for Additional Information for the South Texas Project License Renewal Application Aging Management Program, Set 12 (TAC Nos. ME4936 and ME4937)" (NOC-AE-12002797) (ML12069A024)

RAI 3.1.1.80-1a Editorial Corrections.

In two places, the STPNOC response in enclosure 1 refers to "Table 4 6" of MRP-227-A.

The response should refer to "Table 4-6."

In the response to part 2, the term "lower internal support lower support forging" is used as a reactor internals component identified in MRP-227-A, Table 4-6. The term should have referred to "lower internals assembly lower support forging".

The revised response is provided below. Revisions are shown by change bars in the right-hand margin of the page.

Background:

By letter dated November 21, 2011, the applicant responded to RAI 3.1.1.80-1 that addresses the need for AMR line items to manage cracking or loss of material of reactor vessel internal components using the ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD Program. In its response, the applicant revised LRA Table 3.1.2-1 and Section 3.1.2.2.12. The applicant's revisions indicate that consistent with MRP-227, Revision 0, the PWR Reactor Internals Program is not an applicable aging management program for managing cracking of the components listed in the revised LRA Section 3.1.2.2.12. One of these components listed in the revised LRA Section 3.1.2.2.12 is the upper core support plate. The applicant also indicated that cracking of the upper core plate is managed by the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program.

Sections 3.2.2 and 4.1.1 of the staffs safety evaluation (June 22, 2011; ADAMS Accession No. ML111600498) of MRP-227, Revision 0, address Topical Report Condition 1 for high consequence components. This condition specifies the upper core plate and lower support forging or casting as the expansion components linked to the control rod guide tube (CRGT) assembly lower flange welds, which are the primary components. Section 3.2.2 of the staffs safety evaluation also indicates that inspections of these high consequence components shall be triggered by the degradation of the primary component (in this case, CRGT lower flanges). The staffs safety evaluation further indicates that the examination method for these additional inspections shall be consistent with the examination method used to detect the degradation of the primary component (in this case, EVT-1).

NOC-AE-12002816 Page 2 of 5 Issue:

The staff needs clarification as to whether the PWR Reactor Internals Program identifies the upper core plate as an expansion component linked to the CRGT lower flange welds to manage loss of material due to wear and cracking due to fatigue, as specified in the staffs safety evaluation (June 22, 2011) of MRP-227, Revision 0.

The staff also needs clarification as to whether the applicant's PWR Reactor Internals Program identifies lower internals assembly lower support forging or casting as an expansion component linked to the CRGT lower flange welds to ensure adequate aging management and structural integrity, consistent with the staffs safety evaluation (June 22, 2011) of MRP-227, Revision 0.

Request:

1. Clarify whether the applicant's PWR Reactor Internals Program identifies the upper core plate as an expansion component linked to the CRGT lower flange welds to manage loss of material due to wear and cracking due to fatigue, consistent with the staffs safety evaluation (June 22, 2011) of MRP-227, Revision 0.

2. Clarify whether the applicant's PWR Reactor Internals Program identifies the lower internals assembly lower support forging or casting as an expansion component linked to the CRGT lower flange welds to ensure adequate aging management and structural integrity, consistent with the staffs safety evaluation (June 22, 2011) of MRP-227, Revision 0.

3. Revise the LRA consistent with the applicant's response.

STPNOC Response:

The STP PWR Reactor Internals Program (B2.1.35) was initially prepared using EPRI 1016596, Material Reliability Program: PWR Internals Inspection and Evaluation Guidelines (MRP-227).

The NRC issued Revision 1 of the NRC Safety Evaluation (ML 1308A770) for MRP-227, Rev. 0 on December 16, 2011 and the industry published EPRI-1022863 (MRP-227-A) as an NRC Topical Report in December 2011. As part of this RAI response, the LRA reactor vessel internals sections listed below are revised to be consistent with NRC Safety Evaluation, Revision 1 and MRP-227-A. Enclosure 2 provides the line-in/line-out revision to the following LRA sections:

LRA Table 3.1.1 items 30 and 37 LRA Table 3.1.2-1 LRA Sections 3.1.2.2.6, 3.1.2.2.9, 3.1.2.2.12, 3.1.2.2.15, and 3.1.2.2.17 LRA Appendix A1.35 LRA Table A4-1 Item 27 LRA Appendix B2.1.35

1) NRC Safety Evaluation, Revision 1, Sections 3.2.2 and 4.1.1 and MRP-227-A, Table 4-6 identify the upper core plate as an expansion component linked to the control rod guide tube (CRGT) assembly lower flange welds, which are the primary components. LRA Appendix B2.1.35 and LRA Basis Document, PWRRI (B2.1.35), PWR Reactor Internals program, are revised to include the upper core plate as an expansion component to manage loss of NOC-AE-12002816 Page 3 of 5 material due to wear and cracking due to fatigue, consistent with NRC Safety Evaluation, Revision 1, and MRP-227-A, Table 4-6.

LRA Table 3.1.2-1 is revised to manage the aging effects of cracking due to fatigue in the upper core plate with Aging Management Program (AMP) Water Chemistry (B2.1.2) and PWR Reactor Internals Program (B2.1.35). LRA Table 3.1.2-1 is revised to manage loss of material in the upper core plate due to wear using AMP PWR Reactor Internals Program (B2.1.35). LRA Section 3.1.2.2.12 is also revised to delete the upper core support upper core plate from the list of components not included in the PWR Reactor Internals Program.

2) NRC Safety Evaluation, Revision 1, Sections 3.2.2 and 4.1.1 and MRP-227-A, Table 4-6 identify the lower internals assembly lower support forging as an expansion component linked to the CRGT assembly lower flange welds. LRA Appendix B2.1.35 and LRA Basis Document PWRRI (B2.1.35), PWR Reactor Internals program, are revised to include the lower internals assembly lower support forging as an expansion component linked to the CRGT lower flange welds to manage cracking due to fatigue, consistent with NRC Safety Evaluation, Revision 1 and MRP-227-A,Table 4-6.

An Aging Management Review (AMR) line exists in LRA table 3.1.2-1 to manage cracking of the lower support forging.

3) Enclosure 2 provides the line-in/line-out revision for the changes identified in response to 1 and 2 above.

RAI 3.1.1.80-1a Editorial Correction.

Appendix A1.35 refers to MRP-227-A as "Rev 0". Appendix A1.35, "PWR Reactor Internals," is revised to delete the reference to "Rev. 0". Enclosure 2 provides the line-in/line-out revision to Appendix A1.35.

RAI B1.39-1

1. Editorial Correction In the description of aging management program element 4, Detection of Aging Effects, for Appendix B2.1.39, "Protective Coating Monitoring and Maintenance Program", the AMP states that additional inspections may be necessary depending on "condition assessment" results instead of depending on "inspection" results as was described in the previous version of this AMP. The AMP is revised to reflect that the word "inspection" had been deleted.

The second paragraph under "Program Description" is revised for consistency.

NOC-AE-12002816 Page 4 of 5

2. Supplemental Information Deletion In the description of aging management program element 4, Detection of Aging Effects, for Appendix B2.1.39, the following description is deleted from the program because it is redundant.

The Coating Inspector is certified as a NACE Level II Coating Inspector in accordance with ASTM D 5498. These qualifications are as specified in ASTM D 5163-08.

The Nuclear Coating Specialist or Coating Planner performs coating inspections and both are NACE qualified. This description is in aging management program element 4, Detection of Aging Effects.

3. Supplemental Information Addition In the description of aging management program element 6, Acceptance Criteria, for Appendix B2.1.39, the following information is added to the discussion on "blistering".

All blistering is documented and repaired in accordance with plant procedures.

4. Supplemental Information Deletion In the description of aging management program element 6, Acceptance Criteria, for Appendix B2.1.39, the following information is deleted from the discussion on "flaking/peeling/de-lamination".

If the sum total of the repair area exceeds 25 percent of that item's total painted area or if each individual repair areas exceeds 30 square inches, the condition is documented on a separate process record form,

5. Editorial Correction In the description of aging management program element 10, Operating Experience, for Appendix B2.1.39, in the fifth line of the fourth paragraph, the word "next" appears before the words "Coatings Condition Assessment Walkdown". The word "next" is deleted because the walkdown was performed during the refueling outage 1RE16 noted.

6. Editorial Correction In the description of the Enhancements for Appendix B2.1.39, in the seventh paragraph a "period" inadvertently appears after "D5498". The period is deleted.

NOC-AE-12002816 Page 5 of 5

7. Editorial Correction In the description of the Conclusions for Appendix B2.1.39, the word "enhancements" is revised to singular "enhancement". provides the corrected line-in/line-out revision to Appendix B2.1.39.

Other typographical errors:

The following are typographical errors of information that is "deleted" from the LRA and depicted as line-outs in Enclosures 2 and 3 of Reference 3.

1. On page 31 of 63 of Enclosure 2 in the second row for component type "ductwork" under "NUREG-1801 Vol. 2 Item", the reference "VII.J-6" is deleted. The reference deleted should have been "VII.J-15."

2. On page 36 of 63 of Enclosure 2, Standard Note A is deleted. The information deleted on the page is described as follows:

"Consistent with NUREG-1801 for component, material, environment, and aging effect. AMP is consistent with NUREG-1 802 AMP".

The note description deleted should have been:

"Consistent with NUREG-1801 item for component, material, environment, and aging effect.

AMP is consistent with NUREG-1801 AMP".

3. On page 1 of 2 of Enclosure 3, for item #27 under "Implementation Schedule", EPRI "1016563" is deleted. The EPRI document deleted should have been "1016596."

Chanqe to correspondence reference 4 Reference 4:

STPNOC letter dated November 30, 2011, from D. W. Rencurrel to NRC Document Control Desk, "Annual Update to the South Texas Project License Renewal Application (TAC Nos. ME4936 and ME4937)" (NOC-AE-1 1002758)

(ML11335A140)

Description of Change On page 67 of 78 in Enclosure 2 to the referenced letter, the environment for two solenoid valve component line items in LRA Table 3.3.2-19 are revised. provides the corrected line-in/line-out revision to Table 3.3.2-19.

NOC-AE-12002816 STPNOC LRA Changes with Line-inlLine-out Annotations NOC-AE-1 2002816 Page 1 of 9 List of Revised LRA Sections RAI 3.1.1.80-1a B2.1.39-1 (1)

Affected LRA Section Appendix A1.35 Appendix B2.1.39 Table 3.3.2-19 (1) Not associated with an RAI. Part of the Annual Update letter.

NOC-AE-12002816 Page 2 of 9 A1.35 PWR REACTOR INTERNALS The PWR Reactor Internals program manages cracking, loss of material, loss of fracture toughness, dimensional changes, and loss of preload for reactor vessel components that provide a core structural support intended function. The program implements the guidance of EPRI 1022863, PWR Internals Inspection and Evaluation Guideline (MRP-227-A,-Rev-") and EPRI 1016609, Inspection Standard for PWR Internals (MRP-228). The program manages aging consistent with the inspection guidance for Westinghouse designated primary components in Table 4-3 of MRP-227-A and Westinghouse designated expansion components in Table 4-6 of MRP-227-A. The expansion components are specified to expand the primary component sample should the indications of the sample be more severe than anticipated. The aging effects of a third set of MRP-227-A internals locations are deemed to be adequately managed by existing program components whose aging is managed consistent with ASME Section Xl Table IWB-2500-1, Examination Category B-N-3.

Program examination methods include visual examination (VT-3), enhanced visual examination (EVT-1), volumetric examination, and physical measurements. The program provides both examination acceptance criteria for conditions detected as a result of monitoring the primary components, as well as criteria for expanding examinations to the expansion components when warranted by the level of degradation detected in the primary components. Based on the identified aging effect, and supplemental examinations if required, the disposition process results in an evaluation and determination of whether to accept the condition until the next examination or implement corrective actions. Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection.

The PWR Reactor Internals program is a new program and will be implemented within 24 months after the issuance of EPRI 1022863, PWR Internals Inspection and Evaluation Guideline MRP-227-A.

NOC-AE-12002816 Page 3 of 9 B2.1.39 Protective Coating Monitoring and Maintenance Program Program Description The Protective Coating Monitoring and Maintenance Program manages loss of coating integrity for Service Level 1 coatings inside containment so that the intended functions of post-accident safety systems that rely on water recycled through the containment sump/drain system are maintained consistent with the current licensing basis. The program includes a visual examination of all reasonably accessible Service Level 1 coatings inside containment, including those applied to the steel containment liner, structural steel, supports, penetrations, uninsulated equipment, and concrete walls and floors receiving epoxy surface systems. This program does not include coating of surfaces that are insulated or otherwise enclosed in normal service and concrete receiving a non-film forming clear sealer coat only. This program is consistent with the standards provided in ASTM D 5163-08 and Regulatory Guide (RG) 1.54, Rev. 2, as addressed in NUREG 1801, Rev. 2, XI.S8.

General visual inspections of the containment building Service Level 1 coatings are conducted during every refueling outage. Additional inspections may be necessary depending on iispeetien condition assessment results. Thorough visual inspections are performed on previously designated areas and on areas noted as deficient during the inspection. Characterization of blistering, cracking, flaking, peeling, de-lamination, rusting, and physical damage is performed to allow evaluation of the deficiency for repair, prioritization of repairs, or for future surveillance. Physical testing may be performed when directed by the Nuclear Coating Specialist. Physical tests are performed by individuals trained in applicable referenced standards of Guide D5498. Examinations are conducted by qualified personnel.

Service Level I coatings are not credited for managing loss of material of the steel containment liner.

Aging Management Program Elements The results of an evaluation of each element against the 10 elements described in Appendix A of NUREG-1 800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants are provided below.

Scope of Program (Element 1)

The Protective Coating Monitoring and Maintenance Program includes a visual examination of all reasonably accessible Service Level 1 coatings inside containment, as defined in RG 1.54, Rev. 2. This scope includes coatings applied to the steel containment liner, structural steel, supports, penetrations, uninsulated equipment, and concrete walls and floors receiving epoxy surface systems. This program pertains to the containment interior and equipment, structures or components which are permanently located inside containment. This program does not NOC-AE-12002816 Page 4 of 9 include coating of surfaces that are insulated or otherwise enclosed in normal service and concrete receiving a non-film forming clear sealer coat only.

Service Level I coatings are not credited for preventing loss of material due to corrosion for the steel containment liner.

Preventive Actions (Element 2)

The Protective Coating Monitoring and Maintenance Program does not prevent aging effects but provides measures for monitoring to detect aging prior to loss of intended function.

Coatings are not credited for preventing loss of material.

Parameters Monitored or Inspected (Element 3)

The Protective Coating Monitoring and Maintenance Program inspects coated surfaces for any visible defects, such as blistering, cracking, flaking, peeling, rusting, and physical damage, as specified in ASTM D 5163-08. Any areas of coating discoloration or areas where corrosion has formed under the coating system are documented and evaluated.

Detection of Aging Effects (Element 4)

The South Texas Project (STP) conducts condition assessments of Service Level 1 coatings inside containment during every refueling outage, as specified in ASTM D 5163-08. Additional inspections may be necessary depending on ispeetien condition assessment results.

The Coatings Engineer in charge of the safety-related coatings program meets the qualification criteria for a Nuclear Coatings Specialist in accordance with ASTM D 7108-05.

The Coating Planner is responsible for planning all coating activities, providing technical support to the applicator, and conducting assessment inspections and physical tests when directed by the Nuclear Coating Specialist. The Coating Planner meets the qualification criteria for a Nuclear Coatings Specialist in accordance with ASTM D 7108-05, is a NACE Certified Inspector, and is trained in the applicable referenced standards of Guide D 5498.

These qualifications are as specified in ASTM D 5163-08. The Coating Inspeco*r iS certified as a NACE Level Coating Inspector in accordance with ASTM D 5498. These qualifications arc as specified i ASTM D 5163-08.

The coatings condition assessment includes a visual examination of all accessible Service Level 1 coatings inside containment, including areas near sumps associated with the emergency core cooling system. Thorough visual inspections are conducted to identify and evaluate all accessible areas of degraded coatings, as specified in ASTM D 5163-08, Section 10.1. Location maps and checklists are used to identify the areas to be inspected and to document the inspection results. The coatings condition assessment inspectors have available to them the necessary tools to conduct a thorough inspection of accessible Service Level 1 coatings in containment, consistent with the recommendations in ASTM D 5163-08, Section 10.5.

NOC-AE-12002816 Page 5 of 9 Monitoring and Trending (Element 5)

Prior to performing the inspection, the inspector reviews the two previous coating condition assessment reports. The inspection reports prioritize repair areas as either needing repair during the same outage, needing repair during the next available outage, or monitored and re-evaluated in the next available outage. These monitoring and trending activities are as specified in ASTM D 5163-08.

Acceptance Criteria (Element 6)

As specified in ASTM D 5163-08, paragraph 11, potentially defective coating surfaces identified during the course of an inspection are documented, their severity is evaluated, and corrective actions are taken to ensure there is no loss of intended functions between the inspections. Defective or deficient coating surfaces are prioritized as either needing repair during the same outage, needing repair during the next available outage, or monitored and re-evaluated in the next available outage. The evaluation covers blistering, cracking, flaking, peeling, de-lamination, and rusting as specified below. These acceptance criteria are consistent with, or more conservative than, the acceptance criteria specified in ASTM D 5163-08, subparagraphs 10.2, 10.3, and 10.4.

Blistering-Blistering of any size is a rejectable condition. All blistering is documented and repaired in accordance with plant procedures.

Cracking-Cracking of any size is a rejectable condition. All cracks under 30 mils in width are documented and repaired in accordance with plant procedures. Cracks exceeding 30 mils in width and all cracks associated with delamination are evaluated under the site corrective action program.

Flaking/Peeling/De-lamination-Flaking/peeling/de-lamination of any size is a rejectable condition. All flaking/peeling/de-lamination is documented and repaired in accordance with plant procedures. If them su

,m total of the rep loa oceods 25 percent of that item,'s total paine a;rea or ifec ndvdaepair areas excGeeds 30 square inches, the condition is documnented On a separate process record form Rusting-Comparison with pictorial standards are performed by individuals trained in applicable referenced standards of Guide D5498 on an as-needed basis as determined by the Nuclear Coatings Specialist. The source and extent of rusting is evaluated during the visual examination by the Nuclear Coatings Specialist.

If no defects are found, mark "Coating Intact, No Defects" on the coating condition assessment report form.

NOC-AE-12002816 Page 6 of 9 If portions of the coating cannot be inspected, note the specific areas on the coating condition assessment report form, along with the reason why the inspection cannot be conducted.

Written or photographic documentation, or both, of coating inspection areas, failures, and defects are included in the final coating condition assessment report.

For coating surfaces determined to be suspect, defective, or deficient, destructive/non-destructive tests are performed by individuals trained in applicable referenced standards of Guide D5498 on an as-needed basis as determined by the Nuclear Coatings Specialist.

Corrective Actions (Element 7)

STP site Quality Assurance (QA) procedures, review and approval process, and administrative controls are implemented in accordance with the requirements of 10 CFR 50 Appendix B and are acceptable in addressing corrective actions. The QA program includes elements of corrective action, and is applicable to the safety-related and nonsafety-related systems, structures and components that are subject to aging management review.

Confirmation Process (Element 8)

STP site QA procedures, review and approval process, and administrative controls are implemented in accordance with the requirements of 10 CFR 50 Appendix B and are acceptable in addressing confirmation processes and administrative controls. The QA program includes elements of corrective action, and is applicable to the safety-related and nonsafety-related systems, structures and components that are subject to aging management review.

Administrative Controls (Element 9)

STP site QA procedures, review and approval process, and administrative controls are implemented in accordance with the requirements of 10 CFR 50 Appendix B and are acceptable in addressing confirmation processes and administrative controls. The QA program includes elements of corrective action, and is applicable to the safety-related and nonsafety-related systems, structures and components that are subject to aging management review.

Operating Experience (Element 10)

The South Texas Project conducts condition assessments of Service Level 1 coatings inside containment during every refueling outage. Service Level 1 coatings are inspected during Coating Condition Assessment walkdowns, IWE inspections, Structures Monitoring Program inspections, and through STP's Condition Reporting Process for identification and timely correction of an existing degraded coating condition. A review of Service Level 1 Coatings inspection and repair documentation shows that coating failures identified in Unit 1 and Unit 2 Reactor Containment Building have not been significant. Historically, Service Level 1 coating failures include: mechanical damage, minor isolated cracking measuring less than 30 mils in NOC-AE-12002816 Page 7 of 9 width, and minor surface rusting. Peeling, blistering, and delamination of Service Level 1 coatings that have the potential to block sumps and strainers have not been reported.

In 1992, cracks were identified in the concrete coating on the Unit 1 RCB knockout block wall.

The coating degradation was characterized as a minor crack less than 30 mils in width, not associated with delamination. The degraded coatings were repaired in accordance with the safety-related coatings specification.

In April, 2000, minor surface corrosion on the Unit 2 liner plate at the interface of the liner plate and concrete basemat was identified through the Condition Reporting Process. Coating degradation is characterized as minor rusting. Repairs to degraded coatings were made in accordance with the safety-related coating specification.

In May 2000, an indication approximately 4" x 8" on the Unit 1 containment liner plate was identified near the reactor vessel head lift rig. Engineering investigated and determined that the outer coating was removed with the primed surface below exposed with no signs of corrosion or further coating deterioration noted. The condition was found acceptable as-is.

The indication was re-evaluated in 1 RE16 during the ne*d Coatings Condition Assessment Walkdown. The indication was identified to be approximately the same size and color as was identified in May 2000. The indication shows no signs of corrosion and no streaks of rust on the liner plate below. The size of the indication or its condition has not changed since May 2000; however, the indication will be monitored and re-evaluated during the next outage.

In November 2009, surface corrosion on a hanger support was identified in Unit 1 during the Coatings Condition Assessment Walkdown. The coatings degradation was characterized as minor surface rusting due to condensation. Repairs to degraded coatings were made in accordance with the safety-related coatings specification.

STP has implemented controls for the procurement, application, and maintenance of Service Level 1 protective coatings used inside containment in a manner consistent with the licensing basis and regulatory requirements applicable to the South Texas Project. The requirements of 10 CFR 50 Appendix B are implemented through specification of appropriate technical and quality requirements for the Service Level 1 coatings program which includes ongoing maintenance activities.

Service Level 1 coatings have been tested, selected, and applied to assure that they will withstand nuclear, chemical, and physical conditions of a Design Basis Accident. The historical Service Level 1 coating performance provides reasonable assurance that coating aging effects are managed such that the operability of the Emergency Core Cooling System and the Containment Spray System will not be impaired due to Service Level 1 coating failure.

NOC-AE-12002816 Page 8 of 9 Enhancements Prior to the period of extended operation, the following enhancements will be implemented in the following program elements:

Parameters Monitored or Inspected - Element 3 Procedures will be enhanced to specify parameters monitored or inspected to include; any visible defects, such as blistering, cracking, flaking, peeling, rusting, and physical damage, as specified in ASTM D 5163-08.

Detection of Aging Effects - Element 4 Procedures will be enhanced to specify inspection frequencies, personnel qualifications, inspection plans, inspection methods, and inspection equipment that meet the requirements of ASTM D 5163-08.

Monitoring and Trending - Element 5 Procedures will be enhanced to specify a pre-inspection review of the previous two monitoring reports and, based on inspection report results, prioritize repair areas as either needing repair during the same outage, needing repair during the next available outage, or monitored and re-evaluated in next available outage.

Acceptance Criteria - Element 6 Procedures will be enhanced to include a standardized coating condition assessment report form that will include the identification of coatings found intact with no defects identified, and the identification of coatings that were not inspected and the reason why the inspection cannot be conducted.

Procedures will be enhanced to include a standardized coating condition assessment report that will include written and/or photographic documentation of coating inspection areas, failures, and defects.

Procedures will be enhanced to specify that destructive/non-destructive tests are performed by individuals trained in applicable referenced standards of Guide D5498 on an as-needed basis as determined by the Nuclear Coatings Specialist.

Conclusion The continued implementation of the Protective Coating Monitoring and Maintenance Program, following enhancements, provides reasonable assurance that aging effects will be managed such that the systems and components within the scope of this program will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

NOC-AE-12002816 Page 9 of 9

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Solenoid Valve Solenoid Valve Strainer PB Copper Alloy (IL/oZinc)

TB Copper Allo-y

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Plant Indoor Air Loss of material (x)(Int)

Slan Inndoorr Air None

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Closed Cycle Loss of material Cooling Water (Int)_

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B Closed-Cycle Cooling Water System (B2. 1.10)