NOC-AE-12002855, Response to Request for Additional Information (RAI) 3.2.2.1-1a for the South Texas Project License Renewal Application (TAC ME4936 and ME4937)

Response to Request for Additional Information (RAI) 3.2.2.1-1a for the South Texas Project License Renewal Application (TAC ME4936 and ME4937)

ML12139A131
Person / Time
Site:	South Texas
Issue date:	05/14/2012
From:	Rencurrel D South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-12002855, TAC ME4936, TAC ME4937
Download: ML12139A131 (10)

Text

Nuclear Operating Company South Texas Project ElectricGeneratingStation P.O Box 289 Wadsworth. Texas 77483 *AA -

May 14, 2012 NOC-AE-12002855 10 CFR 54 STI: 33547424 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 Response to Requests for Additional Information (RAI) 3.2.2.1-1a for the South Texas Project License Renewal Application (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 May 14, 2012, "Requests for Additional Information for the Review of the South Texas Project, Units 1 and 2, License Renewal Application

-Aging Management, Set 18 (TAC Nos. ME4936 and ME4937)"

(ML12124A094)

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, STPNOC received RAI 3.2.2.1-1a. The response is provided in Enclosure 1 to this letter.

One new regulatory commitment is added to Table A4-1 of the LRA and is provided in to this letter. 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 Date W. Renr rel Chief Nuclear Officer KJT

Enclosures:

1. STPNOC Response to RAI 3.2.2.1-1a

2. New Regulatory Commitment A~7

NOC-AE-12002855 Page 2 cc:

(paper copy) (electronic copy)

Regional Administrator, Region IV A. H. Gutterman, Esquire U. S. Nuclear Regulatory Commission Kathryn M. Sutton, Esquire 1600 East Lamar Boulevard Morgan, Lewis & Bockius, LLP Arlington, Texas 76011-4511 Balwant K. Singal John Ragan Senior Project Manager Chris O'Hara U.S. Nuclear Regulatory Commission Jim von Suskil One White Flint North (MS 8B1) NRG South Texas LP 11555 Rockville Pike Rockville, MD 20852 Senior Resident Inspector Kevin Polio U. S. Nuclear Regulatory Commission Richard Pena P. O. Box 289, Mail Code: MN116 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)

Washington, DC 20555-0001 Tam Tran Balwant K. Singal License Renewal Project Manager John W. Daily (Environmental) Tam Tran U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission One White Flint North (MS O11F01)

Washington, DC 20555-0001

NOC-AE-12002855 Enclosure 1 Enclosure I STPNOC Response to RAI 3.2.2.1-1a

Enclosure 1 NOC-AE-1 2002855 Page 1 of 5 SOUTH TEXAS PROJECT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION AGING MANAGEMENT, SET 18 (TAC NOS. ME4936 AND ME4937)

Instances of Recurrent Reactor Coolant Leakage (019)

NRC RAI 3.2.2.1-1a

Background:

By letter dated September 22, 2011, the staff issued RAI 3.2.2.1-1 requesting that the applicant state why the stainless steel closure bolting exposed to borated water leakage in LRA Table 3.2.2-4, Safety Injection System, was not managed for cracking. In its response dated November 21, 2011, the applicant stated that the bolting was in an ambient temperature less than 140°F; therefore, stress corrosion cracking is not an applicable aging effect.

NRC Inspection Report No. 05000449/2011005 dated February 13, 2012 describes safety injection (SI) system hot leg check valve 1N122XSI001OA, on which a seal cap enclosure had been installed in 1997 due to reactor coolant leakage from the body to bonnet gasket. The enclosure surrounds the valve bolting, preventing direct inspection. The NRC inspection report also cites multiple occasions of leakage from the seal cap enclosure from 1997 to 2011, indicating that the leak from the bolted connection inside the enclosure was ongoing.

Another instance of borated water leakage is described in the "operating experience" program element of LRA Section B2.1.4, "Boric Acid Corrosion," which states that recurring coolant leakage in reactor coolant pump 2C resulted in the replacement of seal housing bolts in 2004 and 2009. The LRA also states that disassembly of this pump was scheduled for Refueling Outage 2RE15 in October 2011 to perform flatness checks of the seal.

Issue:

1. Given the multiple instances of reactor coolant leakage from the seal cap enclosure of SI system hot leg check valve 1N122XSIOO1OA, the stainless steel bolting within the enclosure may be submerged in a reactor coolant environment with a temperature greater than 140 0 F. The staff noted that the environment within the enclosure may not be sufficiently controlled for oxygen and other contaminants to preclude stress corrosion cracking. The LRA does not contain an AMR item to address the submerged environment and the potential for stress corrosion cracking. It is unclear to the staff how bolting within this and other seal cap enclosures will be age managed, since direct inspection is not possible.

2. Regarding the recurring borated water leakage from reactor coolant pump 2C, the staff does not have sufficient information to determine if the Boric Acid Corrosion program is effective at preventing recurring leakages.

Enclosure 1 NOC-AE-12002855 Page 2 of 5 Request:

1. For all instances where pressure-retaining bolting is surrounded by seal cap enclosures:

a. Describe the bolting alloy and the leaking water environment (i.e., reactor coolant, secondary water).

b. Add aging management review (AMR) items for the aging management of the bolting for loss of material, loss of preload, and cracking due to stress corrosion cracking, as appropriate, in the submerged environment. Provide technical justification for any cases where cracking due to stress corrosion cracking is not included as an applicable aging effect.

c. If the aging management approach in item (b) does not include direct inspection of the bolting, provide technical justification for how the aging effects will be effectively managed during the period of extended operation.

2. Describe the results and any identified corrective actions from the reactor coolant pump 2C seal flatness checks to demonstrate the effectiveness of the Boric Acid Corrosion program.

STPNOC Response:

la. The bolting alloy for SI system hot leg check valve 1N122XSI001OA and the other instances where pressure-retaining bolting is surrounded by seal cap enclosures is A-286 (SA-453 Gr. 660). The leaking water environment is reactor coolant. The material is an iron-based and precipitation hardened high strength material. This material has a high chrome and nickel content and is specifically designed to be resistant to boric acid corrosion. The systems where seal cap enclosures currently exist are the Safety Injection System and the Chemical and Volume Control System. Seal cap enclosures are currently installed on Safety Injection System Check Valve SI0010A in Unit 2 and on Chemical Volume Control System Check Valves CV0001, CV0002, CV0004, and CV0005 in both Unit 1 and Unit 2.The LRA Aging Management Review (AMR) tables that show the material and environments for the affected systems and components are provided below.

Enclosure 1 NOC-AE-1 2002855 Page 3 of 5 Table 3.2.2-4 Engineered Safety Features - Summary of Aging Management Evaluation - Safety Injection System Component Intended Material Environment Aging Effect Aging NUREG- Table I Notes Type Function Requiring Management 1801 Item Management Program Vol. 2 Item Closure LBS, PB, Carbon Borated Water Loss of material Boric Acid V.E-2 3.2.1.45 A Bolting SIA Steel Leakage (Ext) Corrosion (B2.1.4)

Closure LBS, PB, Carbon Plant Indoor Loss of material Bolting Integrity V.E-4 3.2.1.23 B Bolting SIA Steel Air (Ext) (B2.1.7)

Closure LBS, PB, Carbon Plant Indoor Loss of preload Bolting Integrity V.E-5 3.2.1.24 B Bolting SIA Steel Air (Ext) (B2.1.7)

Valve PB Stainless Reactor Cracking ASME Section Xl IV.C2-5 3.1.1.68 A Steel Coolant (Int) Inservice Inspection, Subsections IWB, IWC, and IWD for Class 1 components (B2.1.1) and Water Chemistry (B2.1.2)

Valve PB Stainless Reactor Loss of material Water Chemistry IV.C2-15 3.1.1.83 E, 1 Steel Coolant (Int) (B2.1.2) and One-Time Inspection (B2.1.16)

Enclosure 1 NOC-AE-12002855 Page 4 of 5 Table 3.3.2-19 Auxiliary Systems - Summary of Aging Management Evaluation -

Chemical and Volume Control System Component Intended Material Environment Aging Effect Aging NUREG- Table I Notes Type Function Requiring Management 1801 Vol. Item Management Program 2 Item Closure LBS, SIA Carbon Borated Water Loss of material Boric Acid VII.I-2 3.3.1.89 A Bolting Steel Leakage (Ext) Corrosion (B2.1.4)

Closure LBS, SIA Carbon Plant Indoor Air Loss of material Bolting Integrity VII.l-4 3.3.1.43 B Bolting Steel (Ext) (B2.1.7)

Closure LBS, SIA Carbon Plant Indoor Air Loss of preload Bolting Integrity VII.I-5 3.3.1.45 B Bolting Steel (Ext) (B2.1.7)

Valve PB Stainless Reactor Cracking ASME Section XI IV.C2-5 3.1.1.68 A Steel Coolant (Int) Inservice Inspection, Subsections IWB, IWC, and IWD for Class 1 components (B2.1.1) and Water Chemistry (B2.1.2)

Valve PB Stainless Reactor Loss of material Water Chemistry IV.C2-15 3.1.1.83 E, 2 Steel Coolant (Int) (B2.1.2) and One-Time Inspection (B2.1.16)

Valve LBS, PB, Stainless Treated Loss of material Water Chemistry VII.El-17 3.3.1.91 E, 2 SIA Steel Borated Water (B2.1.2) and One-(Int) Time Inspection S(B2.1.16)

Valve LBS, PB, Stainless Treated Cracking Water Chemistry VII.E1-20 3.3.1.90 E, 2 SIA Steel Borated Water (B2.1.2) and One-(Int) Time Inspection 1(B2.1.16) 1 b. South Texas Project will permanently remove the seal cap enclosures from the affected components at the next available opportunity. After the seal cap enclosures are removed, the component bolting will be replaced or inspected for intergranular stress corrosion cracking. New LRA Table A4-1 Commitment 43 is provided in Enclosure 2. Therefore, the aging effects and environment for the affected components will be consistent with the AMR table description in the LRA. No changes to the AMR tables are required.

Enclosure 2 provides the line-in of new Commitment 43.

lc. The submerged environment associated with the affected components will no longer exist after the seal cap enclosures are permanently removed. The aging effects of the closure bolting of these components will be effectively managed during the period of extended operation by the Boric Acid Corrosion and Bolting Integrity Aging Management Programs.

Enclosure 1 NOC-AE-12002855 Page 5 of 5

2. Seal housing leakage for reactor coolant pump 2C was initially identified by the Boric Acid Walkdown Program in 2004. The joint was disassembled and the gasket was replaced which stopped the leakage. The seal housing leakage for reactor coolant pump 2C was again identified through the Boric Acid Walkdown Program during the Fall 2008 refueling outage (2RE13). A monitoring plan for the reactor coolant pump 2C seal housing leakage was established and permanent repairs were scheduled for the Fall 2011 refueling outage (2RE15). Following a Unit 2 trip in November 2010, the Boric Acid Walkdown Program inspected the seal housing leakage for reactor coolant pump 2C and identified increased leakage. Following plant pressurization to normal operating temperature and pressure, the leak exceeded predetermined leakage criteria. The plant was depressurized to facilitate corrective action implementation. The corrective actions were replacement of the reactor coolant pump seal housing and gasket.

The apparent cause of the reactor coolant pump 2C seal housing leakage is deformation and distortion of the #1 seal housing caused by pressure cycles and successive retightening of the joint. The deformation and distortion interfered with the fit between the seal housing and the thermal barrier, and between the seal housing and the #1 insert support. Progression of deformation and distortion over time eventually contributed to a leak path between the seal housing and the thermal barrier.

As outlined above, the Boric Acid Walkdown Program effectively identified, evaluated and tracked this joint leakage from initial identification to subsequent correction.

The Boric Acid Walkdown Program has not identified any leakage at the seal housing joint of the other seven Reactor Coolant Pumps.

Enclosure 2 NOC-AE-12002855 Enclosure 2 New Regulatory Commitment

Enclosure 2 NOC-AE-12002855 Page 1 of 1 A4 LICENSE RENEWAL COMMITMENTS Table A4-1 identifies proposed actions committed to by STPNOC for STP Units 1 and 2 in its License Renewal Application. These and other actions are proposed regulatory commitments. This list will be revised, as necessary, in subsequent amendments to reflect changes resulting from NRC questions and STPNOC responses. STPNOC will utilize the STP commitment tracking system to track regulatory commitments. The Condition Report (CR) number in the Implementation Schedule column of the table is for STPNOC tracking purposes and is not part of the amended LRA.

Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule 43 The seal cap enclosures from Unit 2 Safety Iniection System Check Valve SI001OA and from B2.1.7 2012 Refueling Unit 1 and Unit 2 Chemical Volume Control System Check Valves CV0001, CV0002, CV0004, Outage (Unit 1)0 2013 and CV0005 will be permanently removed. After removal of the seal cap enclosures, the Refueling Outage component bolting will be replaced or inspected for intergranular stress corrosion cracking. (Unit 2)

CR 12-21155