RBG-47883, Amendment 1 to License Renewal Application

Amendment 1 to License Renewal Application

ML18214A162
Person / Time
Site:	River Bend
Issue date:	08/02/2018
From:	Maguire W Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RBG-47883
Download: ML18214A162 (15)

Text

,. Entergy.

~ Entergy Operations, Inc.

River Bend Station

>485 U.S. Highway 61 N St FrancIsville. LA 70775 Tel 225-381-4374 William F. Maguire Site Vice President RBG-47883 River Bend Station August 2, 2018 Attn: Document Control Desk U.S _Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

Amendment 1 to License Renewal Application River Bend Station , Unit 1 Docket No. 50-458 License No. NPF-47

References:

1) Entergy Letter: License Renewal Application (RBG-47735 dated May 25, 2017)

Dear Sir or Madam:

In accordance with 10 CFR 54 _21 (b), each year following submittal of the license renewal application (LRA) and at least three months before scheduled completion of the Nuclear Regulatory Commission (NRC) review, an amendment to the renewal application must be submitted that identifies any changes to the current licensing basis that materially affects the content of the LRA including the Updated Final Safety Analysis Report (UFSAR) supplement. In the referenced letter, Entergy Operations, Inc. (Entergy) applied for renewal of the River Bend Station , Unit 1 operating license. The LRA amendment 1 for River Bend Station, Unit 1 is provided in Enclosure 1.

No new commitments have been identified in this letter.

In accordance with 10 CFR 50.91 (b)(1), Entergy is notifying the State of Louisiana and the State of Texas by transmitting a copy of this letter to the designated State Official.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August 2, 2018.

~:::'C/dP lU<)A;j ~

Enclosure 1: Annual Update Amendment - River Bend Station

RBG-47883 Page 2 of 2 cc : (with Enclosure)

U. S. Nuclear Regulatory Commission Attn : Emmanuel Sayoc 11555 Rockville Pike Rockville , MD 20852 cc: (w/o Enclosure)

U.S. Nuclear Regulatory Comm ission Region IV 1600 E. Lamar Blvd.

Arlington , TX 76011 -4511 U.S. Nuclear Regulatory Commission Attn : Ms. Lisa M. Regner, Project Manager 09-0 -14 One White Flint North 11555 Rockville Pike Rockville , MD 20852 NRC Senior Resident Inspector Attn: Mr. Jeff Sowa 5485 U.S. Highway 61 , Ste . NRC St. Francisville, LA 70775 Department of Environmental Quality Office of Environmental Compliance Radiological Emergency Planning and Response Section Ji Young Wiley P.O. Box 4312 Baton Rouge , LA 70821-4312 Publ ic Utility Commission of Texas Attn: PUC Filing Clerk 1701 N. Congress Avenue P. O. Box 13326 Austin , TX 78711 -3326 RBF1-18-0143

RBG-47883 Enclosure 1 Annual Update Amendment River Bend Station, Unit 1 License Renewal Application

RBG-47883 Page 2 of 13 1.0 Introduction In accordance with 10 CFR 54.21 (b) , each year following submittal of the license renewal application (LRA) and at least three months before scheduled completion of the NRC review, an amendment to the LRA must be subm itted that identifies any change to the current licensing basis (CLB) that materially affects the content of the LRA, including the Updated Safety Analysis Report (USAR) supplement.

Amendment 1 is based on review of documents potentially affecting the CLB during the period of April 29, 2016 through March 31 , 2018.

2.0 Summary of Changes The review concluded that certain sections of the LRA are affected by changes to the CLB documents and other related LRA reviews . The table below summarizes the changes listing the affected system (if applicable), a description of the change (including the effect on the LRA), and the affected LRA section .

Affected LRA Sections Affected LRA Change Section EC 66879 EC installs safety-related stainless steel tubing , valve , and sight glass for oil recovery between the compressor and evaporator on the HVAC chilled water skid (HVK-CHL 1A) . These components are subject to aging management review (STAMR) . This change is shown on PIO-Table 3.3.2-14 22-14K; however, this PIO is not an LRA drawing . Stainless steel tubing and valve bodies exposed to lube oil are already included in LRA Table 3.3.2-14, chilled water system . In addition , sight glass with a glass material exposed to lube oil is already included. As a result, the only LRA change is the addition of stainless steel sight glass line items in LRA Table 3.3.2-14.

EC 66880 EC installs safety-related stainless steel tubing , valve , and sight glass for oil recovery between the compressor and evaporator on the HVAC chilled water skid (HVK-CHL 18). These components are STAMR.

This change is shown on PIO-22-14L; however, this PIO is not an LRA Table 3.3.2-14 drawing . Stainless steel tubing and valve bodies exposed to lube oil are already included in LRA Table 3.3.2-14, chilled water system . In addition, sight glass with a glass material exposed to lube oil is already included. As a result, the only LRA change is the addition of stainless steel sight glass line items in LRA Table 3.3.2-14.

RBG -47883 Page 3 of 13 Affected LRA Change Section EC 69142 EC changes the applicable system for multiple components due to operating conditions . In general , component identification (tag) numbers were changed to have a prefix of "SFC" instead of "MWS" with a corresponding change in system code from 659 , Makeup Water, to 602 , Fuel Pool Cooling . No plant modifications were performed by this EC. Consequently, no LRA changes are necessary due to the system code change because they do not constitute a material change to the LRA due to components not physically changed and component aging effects remain managed by appropriate programs . Table 3.3.2-18-26 While reviewing this EC, an error was identified. The material type for filter housings MWS-FLT1 A & B is included in LRA Table 3.3.2-18-26 as carbon steel. Instead, they are made of stainless steel (reference data sheet in EC 69142 reference folder downloaded from internet using model number from AS Parameters). MWS-FLT1 A & B are the only two filter housings included in the LRA table . Thus, LRA Table 3.3.2-18-26 is revised to replace the filter housing line items shown with a material of carbon steel with filter housing line items shown with a material of stainless steel exposed to the same environment. No other LRA changes are necessary as a result of this EC.

Technical change to LRA Table 3.3.2-18-1 While reviewing EC 70706 , it was noted that component types such as sight glass, valve bodies, etc. in the lube oil system associated with pumps C11 -PC001A and B are included in LRA Table 3.3.2-18-1, "Control Rod Drive Hydraulic System , Nonsafety-Related Components Affecting Safety-Related Systems," but piping and tubing exposed to lube oil were not included. In addition , filter housings RDS-FLTD011 A and B are included in LRA Table 3.3.2-18-1 with an environment of treated water > 140°F, which is incorrect. Instead, they are exposed to Table 3.3.2-18-1 lube oil, and no table line item exists with this component type/material/environment combination. Consequently, line items are added to LRA Table 3.3.2-18-1 for carbon steel piping and filter housings and stainless steel tubing exposed to lube oil. Carbon steel filter housings exposed to treated water remain in the system ;

therefore , no change is necessary to existing LRA table line items.

LRA Table 3.3.2-18-1 already includes appropriate external environment line items for piping, tubing , and filter housing component types ; therefore, only three line items addressing the lube oil internal environment are added .

RBG-47883 Page 4 of 13 Affected LRA Change Section Letter RBF1-18-0015 Letter RBF1 0015 includes a request for alternative to use BWRVIP guidelines in lieu of specific ASME code requirements.

Results of the letter do not require a change to the LRA; however, the letter cites ASME Code Section XI , 2007 Edition through 2008 Addenda as the code of record for fourth interval inservice inspections Section A. 1.13 (lSI) . A review of SEP-ISI-RBS-001 , "ASME Section XI , Division 1 Section A.1 .22 RBS Inservice Inspection Program ," confirms the 2007 edition through Section A.1.23 2008 addenda is the applicable ASME Code edition for the fourth Section A.1.36 interval. This letter specifies a later edition of the ASME Code than Section B.1.13 the ASME Code specified in the LRA for lSI inspections. LRA Section B.1.22 Appendix A (USAR supplement) is revised to remove reference to a Section B.1.23 specific ASME Code year as this is not necessary for the USAR Section B.1.36 supplement. In addition , applicable program sections of Appendix A are revised to include a statement that every 10 years programs are updated to the latest ASME Section XI code edition and addendum approved by the NRC in accordance with 10 CFR 50.55a. LRA Appendix B programs are revised to cite ASME Code 2007 Edition through 2008 Addenda.

Technical change to LRA Table 3.3.2-7 Response to RAI3.3.1 -1 was submitted by Letter RBG-47818 dated February 7, 2018. In th is RAI response, several line items were revised in LRA Table 3.3.2-7, "Fire Protection - Water System ," to include flow blockage as an applicable aging effect. Included in these revised line items is an inadvertent change adding flow blockage as Table 3.3.2-7 an applicable aging effect to copper alloy valve bodies exposed externally to raw water. Flow blockage is not an applicable aging effect for valve bodies exposed to an external environment. Thus, LRA Table 3.3.2-7 is revised to remove the aging effect of flow blockage from copper alloy valve bodies exposed externally to raw water.

Technical change to LRA Appendix A, Section A.1 LRA Appendix A, "Updated Safety Analysis Report Supplement,"

Appendix A, Section A.1, is revised to clarify the frequency of performance of aging Section A.1 management program activities. Upon issuance of the renewed license, this summary description will be incorporated into the licensing basis.

RBG-47883 Page 5 of 13 Affected LRA Change Section Technical change to LRA Appendix A and Appendix B During license renewal activities, a discrepancy was found in LRA Appendix A, Section A.1 .17, External Surfaces Monitoring (ESM) .

Specifically, the description of the ESM Program erroneously cites Appendix A, reduction of heat transfer as an applicable aging effect. As described Section A.1 .17 in NUREG-1801 , Revision 2,Section XI.M36 , "External Surfaces Monitoring of Mechanical Components," reduction of heat transfer is Appendix B, not an applicable aging effect for the NUREG-1801 program. In Section B.1.17 addition, there are no components included in the LRA where the aging effect of reduction of heat transfer is managed by the ESM Program . Thus , reduction of heat transfer is removed as an applicable aging effect in LRA Appendix A and LRA Appendix B.

RBG-47835 Entergy responded to RAI 3.1.2.1.2-2 in letter RBG -47835 dated March 26, 2018. In the response to RAI 3.1.2.1.2-2, LRA Table 3.1.2-2 and LRA Table 3.1 .1, Item 3.1.1-103, were revised to remove Inservice Inspection as an applicable program for managing aging effects of the in-core instrument dry tubes and stabilizers. Also Section 3.1.2.1.2 included in the response was an intended change to LRA Section 3.1.2.1 .2, "Reactor Vessel Internals," by removing Inservice Inspection as an applicable program ; however, "Inservice Inspection" was inadvertently not removed as intended . Thus , "Inservice Inspection" is removed from LRA Section 3.1 .2.1.2 in this LRA annual amendment.

Because LRA tables were appropriately revised in the response to RAI 3.1.2.1 .2-2, no other changes are necessary.

RBG-47883 Page 6 of 13 3.0 LRA Changes RBS LRA changes are shown below. Additions are shown with underline and deletions with strikethrouah .

3.1.2.1.2 Reactor Vessel Internals Aging Management Programs The following aging management programs manage the aging effects for the reactor vessel internals components.

• BWR Vessel Internals

• Inservice Inspection

• Water Chemistry Control- BWR Table 3.3.2-7 Fire Protection - Water System Summary of Aging Management Evaluation Table 3.3.2-7: Fire Protection - Water System Aging Effect Aging Component Intended Requiring Management NUREG-1801 Table 1 I Type Function Material Environment Management Program Item Item Notes Valve body Pressure Copper alloy Raw water Loss of material Fire Water VII.G.AP-197 3.3.1-64 B boundary (ext) and flow blockage System

RBG-47883 Page 7 of 13 Table 3.3.2-14 Chilled Water System Summary of Aging Management Evaluation Table 3.3.2-14: Chilled Water System Aging Effect Aging Intended Requiring Management NUREG-1801 Table 1 Component Type Function Material Environment Management Program Item Item Notes Sight glass Pressure Stainless Air - indoor None None VII,J .AP-123 3.3.1-120 6 boundar~ steel (ext)

Sight glass Pressure Stainless Lube oil (int) Loss of material Oil Anal~sis VII,C2.AP-138 3.3.1-100 A, 302 boundar~ steel Table 3.3.2-18-1 Control Rod Drive Hydraulic System Nonsafety-Related Components Affecting Safety-Related Systems Summary of Aging Management Evaluation Table 3.3.2-18-1: Control Rod Drive Hydraulic System, Nonsafety-Related Components Affecting Safety-Related Systems Aging Effect Aging Component Intended Requiring Management NUREG-1801 Table 1 Type Function Material Environment Management Program Item Item Notes Filter housing Pressure Carbon steel Lube oi l (int) Loss of material Oil Anal~sis VII,C1 .AP-127 3.3.1-97 C,302 bounda[Y EiQlng Pressure Carbon steel Lube oil (int) Loss of material Oil Anal~sis VII,C1.AP-127 3.3.1-97 C,302 boundar~

Tubing Pressure Stainless steel Lube oil (int) Loss of material Oil Anal~sis VII,C1 .AP-138 3.3.1-100 C,302 boundar~

RBG-47883 Enclosu re 1 Page 8 of 13 Table 3.3.2-18-26 Makeup Water System Nonsafety-Related Components Affecting Safety-Related Systems Summary of Aging Management Evaluation Table 3.3.2-18-26 Makeup Water System , Nonsafety-Related Components Affecting Safety-Related Systems Aging Effect Aging Intended Requ iring Management NUREG-1801 Table 1 Component Type Function Material Environment Management Program Item Item Notes Filter housing Pressure Gaf99R steel Air - indoor b9ss 9f fflatefial e~d e fRal gl:lffases VILLA 77- ~.~ . ~ 18 A boundary Stainless (ext) None M9RitmiRg VILJ .AP-123 3.3.1-120 steel None Filter housing Pressure Gaf99R steel Treated water Loss of material Water Chemistry }JILG~ . Af2 ~Q~ ~.~.~ 48 G~

boundary Stainless (int) Control - Closed VILC2.A-52 3.3.1-49 steel Treated Water Systems

RBG-47883 Enclosure 1 Page 9 of 13 Appendix A A.l AGING MANAGEMENT PROGRAMS The integrated plant assessment for license renewal identified aging management programs necessary to provide reasonable assurance that structures and components subject to aging management review will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation. This section describes the aging management programs and activities required during the period of extended operation. Aging management programs will be implemented prior to entering the period of extended operation. The specified frequency for each periodic aqing management program activity is met if the activity is performed within 1.25 times the interval specified in the program description, as measured from the previous performance or as measured from the time a specified condition on the frequency is met.

The corrective action, confirmation process, and administrative controls of the RBS (10 CFR Part 50, Appendix B) Quality Assurance Program are applicable to all aging management programs and activities during the period of extended operation. RBS quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B. The RBS Quality Assurance Program applies to safety-related and important to safety structures and components. Corrective actions and administrative (document) control for both safety-related and nonsafety-related structures and components are accomplished in accordance with the established RBS corrective action program and document control program and are applicable to all aging management programs and activities during the period of extended operation. The confirmation process is part of the corrective action program and includes reviews to assure adequacy of corrective actions, tracking and reporting of open corrective actions, and review of corrective action effectiveness. Any follow-up inspection required by the confirmation process is documented in accordance with the corrective action program.

A.1.13 Containment Inservice Inspection -IWE The Containment Inservice Inspection - IWE Program is implemented through plant procedures which provide administrative controls for the conduct of activities that are necessary to fulfill the requirements of 10 CFR 50.55a, which imposes the inservice inspection (lSI) requirements of the ASME B&PV Code,Section XI, Subsection IWE, for steel containments (Class MC) and steel liners for concrete containments (Class CC). There are no tendons associated with the RBS steel containment vessel (SCV). The RBS containment system is a General Electric BWR Mark III pressure suppression containment system consisting of a drywell, vapor suppression pool, and a primary containment structure. The RBS primary containment structure is a low-leakage, free-standing SCV consisting of a vertical upright cylinder with a torispherical dome and a flat liner plate at the base. The SCV forms the containment pressure boundary and encloses the vapor suppression pool and the drywell.

RBG-47883 Page 10 of 13 The program includes the SCV and its integral attachments, containment equipment hatches, airlocks, and pressure-retaining bolting. The program performs visual examinations (general visual , VT-1 and VT-3) to assess the general condition of the containment and to detect evidence of degradation that may affect structural integrity or leak tightness. The visual inspections monitor the condition of the SCV surface areas, including welds and base metal and integral attachments; personnel and equipment access hatches; and pressure-retaining bolting. Bolting is not susceptible to cracking and does not require surface or volumetric examinations to detect cracking per IWE. The Containment Inservice Inspection -IWE program specifies acceptance criteria, corrective actions, supplemental inspections as required, and provisions for expansion of the inspection scope when identified degradation exceeds the acceptance criteria. The code of record for the examination of the RSS Glass MG components and related requirements is in accordance with ASME Code Section XI, Subsections IWE, 2001 Edition with the 2003 Addenda, as mandated and modified by 10 CFR 50.55a. Ever;l1 0 years this program is updated to the latest ASME Section XI code edition and addendum approved by the NRC in accordance with 10 CFR 50.55a.

A.1.17 External Surfaces Monitoring The External Surfaces Monitoring Program manages aging effects of components fabricated from metallic, elastomeric, and polymeric materials through periodic visual inspection of external surfaces for evidence of loss of material, cracking, reduction of hoat transfer, reduced thermal insulation resistance, and change in material properties. When appropriate for the component and material, physical manipulation, such as pressing, flexing and bending, is used to augment visual inspections to confirm the absence of elastomer hardening and loss of strength. The External Surfaces Monitoring Program is also credited for situations where the material and environment combinations are the same for the internal and external surfaces such that the external surfaces are representative of the internal surfaces.

A.1.22 Inservice Inspection The Inservice Inspection (lSI) Program manages cracking , loss of material, and reduction in fracture toughness for ASME Class 1, 2, and 3 pressure-retaining components including welds , pump casings, valve bodies, integral attachments, and pressure-retaining bolting using periodic volumetric, surface, and visual examination and leakage testing as specified in ASME Section XI code, ~

Edition, 2003 addendumas mandated by 10 CFR 50.55a. Additional limitations, modifications, and augmentations described in 10 CFR 50.55a are included as a part of this program. Every 10 years this program is updated to the latest ASME Section XI code edition and addendum approved by the NRC in accordance with 10 CFR 50.55a. Repair and replacement activities for these components are covered in Subsection IWA of the ASME code edition of record.

RBG-47883 Enclosure 1 Page 11 of 13 A.1.23 Inservice Inspection - IWF The Inservice Inspection (l SI) -IWF (ISI-IWF) Program performs periodic visual examinations of ASME Class 1, 2, and 3 piping and component supports to determine general mechanical and structural condition or degradation of component supports. The examinations include verification of clearances, settings and physical displacements, and identification of loose or missing parts, debris, corrosion , wear, erosion , or the loss of integrity at welded or bolted connections. The ISI-IWF Program is implemented through plant procedures which provide administrative controls , including corrective actions, for the conduct of activities that are necessary to fulfill the requirements of ASME Section XI , as mandated by 10 CFR 50.55a. Every 10 years this program is updated to the latest ASME Section XI code edition and addendum approved by the NRC in accordance with 10 CFR 50.55a. The monitoring methods are effective in detecting the applicable aging effects, and the frequency of monitoring provides reasonable assurance that significant degradation can be identified prior to a loss of intended function .

A.1.36 Reactor Head Closure Studs The Reactor Head Closure Studs Program manages cracking and loss of material due to wear or corrosion for reactor head closure sruas bolting (studs, washers , nuts, and flange threads) using inservice inspection (ASME Section XI , 2001 Edition , 2003 Addendum ,

Table I'/I/B 2500 1) and preventive measures to mitigate the effects of aging. Preventive actions include use of an acceptable surface treatment , use of stable lubricants, use of bolting materials with low susceptibility to SCC, and avoidance of the use of metal-plated stud bolting. The program detects cracks , loss of material , and leakage using visual, surface, and volumetric examinations as required by ASME Section XI. Every 10 years this program is updated to the latest ASME Section XI code edition and addendum approved by the NRC in accordance with 10 CFR 50.55a. The program also relies on recommendations to address reactor head closure bolting degradation listed in NUREG-1339 and NRC RG 1.65.

Appendix B 8.1.13 CONTAINMENT INSERVICE INSPECTION -IWE The program includes the SCV and its integral attachments, containment equipment hatches, airlocks, and pressure-retaining bolting. The program performs visual examinations (general visual , VT-1 and VT-3) to assess the general condition of the containment and to detect evidence of degradation that may affect structural integrity or leak tightness. The visual inspections monitor the condition of the SCV surface areas, including welds and base metal and integral attachments; personnel and equipment access hatches; and pressure-retaining bolting . Bolting is not susceptible to cracking and does not require surface or volumetric examinations to detect cracking per IWE. The CII-IWE Program specifies acceptance criteria , corrective actions, supplemental inspections as required , and provisions for expansion of the inspection scope when identified degradation exceeds the acceptance criteria. The code of record for the examination of the RBS Class MC components and related requirements is ASME Code Section XI , Subsections IWE, 2GG+2007 Edition with the ~2008 Addenda, as mandated and modified by 10 CFR 50.55a. Every 10 years

RBG-47883 Enclosure 1 Page 12 of 13 this program is updated to the latest ASME Code Section XI edition and addendum approved by the NRC in accordance with 10 CFR 50.55a.

B.1.17 EXTERNAL SURFACES MONITORING The External Surfaces Monitoring Program manages aging effects of components fabricated from metallic, elastomeric, and polymeric materials through periodic visual inspection of external surfaces for evidence of loss of material, cracking , reduction of hoat transfer, reduced thermal insulation resistance, and change in material properties. When appropriate for the component and material, physical manipulation, such as pressing, flexing, and bending , is used to augment visual inspections to confirm the absence of elastomer hardening and loss of strength. The External Surfaces Monitoring Program is also credited for situations where the material and environment combinations are the same for the internal and external surfaces such that the external surfaces are representative of the internal surfaces.

B.1.22 INSERVICE INSPECTION The Inservice Inspection (lSI) Program manages cracking, loss of material , and reduction in fracture toughness for ASME Class 1,2, and 3 pressure-retaining components including welds, pump casings, valve bodies, integral attachments, and pressure-retaining bolting using periodic volumetric , surface, and visual examination and leakage testing as specified in ASME Section XI code, 2GG+2007 Edition , 2GGd2008 addendum. Additional limitations, modifications, and augmentations described in 10 CFR 50.55a are included as a part of this program . Every 10 years this program is updated to the latest ASME Section XI code edition and addendum approved by the NRC in accordance with 10 CFR 50.55a. Repair and replacement activities for these components are covered in Subsection IWA of the ASME code edition of record.

B.1.23 INSERVICE INSPECTION - IWF The Inservice Inspection (lSI) - IWF (ISI-IWF) Program performs periodic visual examinations of ASME Class 1, 2, and 3 piping and component supports to determine general mechanical and structural condition or degradation of component supports. The examinations include verification of clearances, settings and physical displacements, and identification of loose or missing parts, debris, corrosion, wear, erosion , or the loss of integrity at welded or bolted connections. RBS MC component supports are addressed under the ASME Section XI , Subsection IWE program. The ISI-IWF Program is implemented through plant procedures which provide administrative controls, including corrective actions, for the conduct of activities that are necessary to fulfill the requirements of ASME Section XI , as mandated by 10 CFR 50.55a. The monitoring methods are effective in detecting the applicable aging effects, and the frequency of monitoring provides reasonable assurance that significant degradation can be identified prior to a loss of intended function.

RBG-47883 Page 13 of 13 RBS is in its tRifafourth 10-year lSI inspection interval. The ISI-IWF Program was developedis implemented in accordance with ASME Section XI, ~2007 Edition through the aQG.32008 Addenda as approved by 10 CFR 50.55a. In accordance with 10 CFR 50.55a(g)( 4)(ii), the program is updated each successive 120-month inspection interval to comply with the requirements of the latest edition of the ASME Code specified twelve months before the start of the inspection interval.

8.1.36 REACTOR HEAD CLOSURE STUDS The Reactor Head Closure Studs Program manages cracking and loss of material due to wear or corrosion for reactor head closure stud bolting (studs, washers, nuts, and flange threads) using inservice inspection (ASME Section XI , ~2007 Edition , aQG.32008 Addendum, Table IWB-2500-1) and preventive measures to mitigate the effects of aging. Preventive actions include use of an acceptable surface treatment, use of stable lubricants, use of bolting materials with low susceptibility to SCC, and avoidance of the use of metal-plated stud bolting. The program detects cracks, loss of material, and leakage using visual , surface, and volumetric examinations as required by ASME Section XI. Every 10 years this program is updated to the latest ASME Section XI code edition and addendum approved by the NRC in accordance with 10 CFR 50.55a. The program also relies on recommendations to address reactor head closure bolting degradation listed in NUREG-1339 and NRC RG 1.65.