Text

Inservice Inspection (ISI) Owner'S Activity Report )OAR) for Refueling Outage 22
	ML16068A174
Person / Time
Site:	Summer
Issue date:	03/03/2016
From:	Lippard G; South Carolina Electric & Gas Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RC-16-0034
	Download: ML16068A174 (23)
	v • d • e

~1 George A. Lippard Vice President, Nuclear Operations 803.345.4810 A SCANA COMPANY March 3, 2016 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

Dear Sir / Madam:

Subect: VIRGIL C. SUMMER NUCLEAR STATION (VCSNS), UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 INSERVICE INSPECTION (181) OWNER'S ACTIVITY REPORT (OAR) FOR REFUELING OUTAGE 22 Attached is the Inservice Inspection (1S1) Owner's Activity Report (OAR) covering ISI activities associated with the second outage for the first period of the fourth ISI interval and the second outage for the third period of the second Containment Inservice Inspection (CISI) interval. This report is submitted pursuant to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (2007 Edition through 2008 Addenda), Section Xl, Rules for Inservice Inspection of Nuclear Power Plant Components, Article IWA-6000, Records and Reports, and ASME Code Case N-532-4, Repair/Replacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission, Section Xl, Division 1.

Should you have any questions, please call Mr. Bruce Thompson at (803) 931-5042.

WLT/GALlwm Attachments:

1. Inservice Inspection Owner's Activity Report for Refuel 22, Report Number 21

2. Containment Inservice Inspection -2015 ASME Section Xl, Subsections IWE and IWL Responsible Engineer Report c: K. B. Marsh S. A. Williams S. A. Byrne NRC Resident Inspector J. B. Archie K. M. Sutton N. S. Camns NSRC J. H. Hamilton RTS (LTD 323)

J. W. Williams File (810.19-3, RR5300)

W. M. Cherry PRSF (RC-16-0034)

C. Haney V.C.Summer Nuclear Station

P.O.Box 88.* Jenkinsville, SC. 29065 ;,F (803) 941-9776 Aoqq

Document Control Desk USNRC, Region II RC-1 6-0034 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS), UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 INSERVICE INSPECTION OWNER'S ACTIVITY REPORT FOR REFUEL 22, REPORT NUMBER 21

FORM OAR-I OWNNER'S ACTIVITY REPORT Report Number INSER VICE INSPECTION REPORT #21.

PatVIRGIL C. SUMMEVR NUCLEAR STATION, P0 BOX 88, JENKINS-VILLE, SC 29065 Unit No. UNIT 1 Commercial service date 1/1/1984 Rfeigutgeo.RF-22 tiFapplicable)

Current inspection interval II-4h W /W n (In!, 2nid. 3rd, 4ti, oilier)

Current inspection period ISI - 1st; IWE/IWL - 3rd (iln, 2nd, 3rd1)

Edition and Addenda of Section Xl applicable to the inspection plans I51 2007-2008a; IVWE/IVVL 2001:-2003a Dat an rvisonofnsecton ISI-March 31, 2014 Revision 0; IWE/IWL-March 4, 2013, Revision 2 Dat rvison an f nspctinplans Edition and Addenda of Section XI-applicable to repair/replacement activities, if different than the inspection pla'nsN/

Code Cases used: N-532-4, N-706-1, N-729-1, N-770-I, N-722- 1 (ifnpplicable)

CERTIFICATE OF CONFORMANCE I certify that (a) the statements made in this report are correct; (b) the examinations end tests meet the Inspection Plan as required by the ASME Code, Section Xl; and (c) the repair/replacement activities and evaluations supporting the completion of 1-F-22 (refelnilng sutego number) conform to th ~ureet.d Section Xl .. .

Owner or Owner's Designee, Title CERTIFICATE OF INSERVICE INSPECTION I, the undersigned, holding a valid commission Issued by the National Board of Boiler end Proesure Vessel Inspoctoreand the State or Province of SOUZh CA*OeUNA end employed by 1isa GLOSAL 5"/TANARD5 of HNITFORiD.C have'inspactad the Items described In this Ownerds Activity Report, and state that, to the beet of my knowledge and belief, the Owner ha-s performed eli activities represented by this report in accordance with the requiremonts of Section XI, By signing this cerilficate neither the Irspector nor his employer makes any warranty, expressed or Implied, concerning the rpairlr replacement activities end evaluation dscribad in this report. Furthermore, neither the Inspector nor his employer shall be liable in any manner for any personal InJur or pro rtry damage or a lose of any kind edising from or connected with this Inspection.

B.MsT**eKO~e */* f ,,t't Commissions NB#1 3930 SC#264 ANi Nooet lrn. tto Poie* e Edgtnntl Qota

VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 OWNER'S ACTIVITY REPORT FOR REFUEL 22 TABLE 1 SITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE EXAMINATION CATEGORY AND ITEM NUBRITEM DESCRIPTION EVALUATION DESCRIPTION N-7291 XRE0001dry boric acid crumb deposit remnants from N-7291 XRE00012006 canopy seal leak B4.10 TOP OF VESSEL HEADCR5-19 F-A SPH-0037 wire debris present at load pin between eye 1.20R RIGID HANGER nut and clevis -CR-i15-05124 no wear or loss of function lower cotter pin on pipe clamp missing one F-A SPH-0120 tail - washer not installed on N side of pipe 1.20R RIGID HANGER clamp - no loss of function CR-i15-05121 F-A SH-0 08outer nut on hanger tie rod not seated - no F-A MS-0108loss of function 1.20SP SPRING HANGER C- -53 F-A CCH-0327 slight rotation of pipe support - no loss of 1.30 RIGD HAGERfunction 1.30 RIGD HAGERCR-i15-01549 F-A EF-5039pipe support clamp found rotated during F-A EF-5039walkdown - realigned 1.20R RIGID HANGER CR-15-04915 F-A CH0941walkdown ID'd spring can not within the cold 1.10SP SPRING HANGER setting - support cold setting adjusted CR-I15-05128 F-A CSH-0971 .jam nut found loose - tightened - no loss of 1.10SP SPRING HANGER function

_____________________________CR-I15-05167 OAR-i Table 1 ag 1 Page

VIRGIL C. SUMMER NUCLEAR STATION UNIT I OWNER'S ACTIVITY REPORT FOR REFUEL 22 TABLE 2 ABSTRACT OF REPAIR/REPLACEMENT ACTIVITIES REQUIRED FOR CONTINUED SERVICE COECASITEM DESCRIPTION DATE REPAIR/REPLACEMENT DESCRIPTION OF WORK COMPLETED PLAN NUMBER IFV03531-EF Flow Control Replaced Plug/Stem 3Valve (Emergency assembly due to seat 11/22/15 1513005

___________Feedwater) leakage. CR-I15-06088 FCV001 13A-CS Boric Acid OverHaul Actuator and 3inlet flow control valve Valve as part of CR 10/23/15 1416715 (Chemical & Volum Control) 04454 XVB031 21 B-SW Diesel 3Generator Cooler B SW Flange Replacement 10/30/15 0712917 Retumn Valve (Service Reference CR-06-03587 Water)_____________ ____________ ____________ __

XVD08318B-CS Boric Acid FilerOuletVave(Chemical Replaced Valve Intemnals 771 575 FilterOutletValveReference CR-I 5-03202 771 575

& Volume Control)

XVD08318A-CS Boric AcidRelcdVventms 3FleOultVle(hmclReference CR-i15-03122 7/17/15 1507264

& Volume Control)

XVA1 6729-SF RMU system Replaced Valve stem and 3to SF supply isolation valve. Machined high spots on 9/21/15 .1416716 (Spent Fuel) valve body CR-i12-00776 3- XVC02876B-MS Check Replaced XVC-02876B-Vav Mi ta) MS Check Valve 11/12/15 1404610 Valv Stam)

(Man Reference CR-1 4-02146 Replaced Trim Kit on

'1 XVT08096A-RC Head vent XVT08096A-RC Head vent' 11/10/15 1417139 Valve (Reactor Coolant) Valve Reference CR 03979 2XVT02843-MS Globe Valve Globe Valve Replacement 11/14/15 1416416 (Main Steam) Reference CR-i14-06071

!IFV355-EFFlowConrolReplaced. Plug/Stem 1Va355veF FlowrContro assembly due to seat1/2155085 2Fale.d(Emergecy leakage.Reference CR-IS 112415-00 Feedwter)00162 XVTO8145-CS RCS PZR Replaced Internals Trim Spray Hdr Isol Valve. Roll Pin,Plug,Stem and 1 ~(Chemical & Volume Cage Reference CR-14-9//i1472

___________Control) 03979 XVCO48OCCSCeck VC-08480C Valve Body Valv (CemialVoumeRepair and Check Valve 1/21 016 2 aleC hemcalrVlum Replacement Reference 1/2i 016 Control)CR-i10-02912 XVMO8O1

-MS SIV XVM-02801B MSIV Valve 2(Main Steam) Body Repair Reference 11/10/15 1511083 CR-I15-05098 Page 1 Page 1OAR-I Table 2

VIRGIL C. SUMMER NUCLEAR STATION UNIT I OWNER'S ACTIVITY REPORT FOR REFUEL 22 TABLE 2 (CONTINUED)

ABSTRACT OF REPAIR/REPLACEMENT ACTIVITIES REQUIRED FOR CONTINUED SERVICE ITEM DESCRIPTION DATE REPAIR/REPLACEMENT COECASDESCRIPTION OF WORK COMPLETED PLAN NUMBER XPP0048C Suction Flange 3 ~~XPP0048C-VU C Chill Studs and Nuts5//11437/462 Water Pump (Chill Water) Replacement. Reference CR-i14-05992 Replaced Valve 3 ~~XVC09682A-CC "A' CC XVC09682A-CC due to it 1/71 333 Pump Disch Valve being worn and needed replacement.____________

XVT08I49A-CS RCS Replaced Internals in Letdown Flow Control Valve. XVT08149A-CS based on 1/21 474 (Chemical & Volume Extent of Condition.

Control)aszx Reference CR-I14-03979 MK-BDH-0013 PSA-1/4 Replaced snubber 2Snubber Reference CR-I15-04856 1/41 423 MK-MSH-0246 PSA-10 Replaced snubber 2Snubber Reference CR-i15-052171/351473 MK-RHH-0167 PSA-10 Replaced snubber 2Snubber Reference CR-15-05497 1/31 457 MK-MSH-1 642 PSA-1 Replaced snubber Snubber Reference CR-15-05319 1/31 4S7 MK-EFH-0180 PSA-1 Replaced snubber Snubber Reference CR-15-04981 1/31 456 MK-CSH-0916 PSA-1/2 Replaced snubber 1Snubber Reference CR-15-05418 1/31 455 RCH-0050/CGE 1-4502-14 Replaced nuts on RCH-1 Pipe Clamp Reactor 0050 Reference CR-IS- 11/18/i5 1500311

_______________Coolant 05521 ____________

3 XW03O14B-SP Vacuum Replaced disc and cage 11/30/15 1502182

_______________Breaker Valve on XW03014B-SP __________ ___________

XHE0002A Service Water Replaced studs on 3 XHE0002A Service Water 10/23/15 1409596 Heat Exchanger Heat Exchanger ____________

Page 2 Page 2OAR-i Table 2

Document Control Desk USNRC, Region II RC-1 6-0034 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS), UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER REPORT

CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL VIRGIL C. SUMMER NUCLEAR STATION RESPONSIBLE ENGINEER EVALUATION REPORT February 18, 2016 Prepared By:

Dale D. Krause, P.E.

IWEIIWL Responsible Engineer

CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT TABLE OF CONTENTS SECTION DESCRIPTION PAGE

1.0 INTRODUCTION

1 2.0 SCOPE 1 3.0 INSPECTION PERSONNEL 2 4.0 IWE/IWL INSPECTIONS 3 5.0 INSPECTION PHILOSOPHY 3 6.0 RESPONSIBLE ENGINEER EVALUATION 4 6.1 IWE Evaluation 4 6.1.1- IWE Augmented Inspections 4 6.1.2 IWE Other Inspection Results 8 6.1.3 IWE Reportable Conditions 8 6.1.4 IWE Reportable Conditions Requiring Augmented 8 Inspections 6.2 IWL Evaluation 8 6.2.1 IWL Augmented Inspections 8 6.2.2 IWL Prestress Tendon System Evaluation 9 6.2.3 IWL Containment Exterior Concrete Surfaces 12 Evaluation 6.2.4 IWL Reportable Conditions 12 6.2.5 IWL Reportable Conditions Requiring Augmented 12 Inspections 7.0

SUMMARY

AND CONCLUSIONS 13 PAGEi

CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT

1.0 INTRODUCTION

This report evaiuates the Containment Inservice Inspections which were conducted in accordance with the requirements of I10CFR50.55a at the Virgil C.

Summer Nuclear Station (VCSNS) during the period of October-November, 2015.

2.0 SCOPE The Containment Inservice Inspection (CISI) Program Plan (ISE-4) details the requirements for the examination and testing of ASME Section XI Class MC and Class CC components at the VC Summer Nuclear Station (VCSNS). This Program Plan was developed in accordance with the requirements of the 2001 Edition (with Addenda through 2003) of the ASME Boiler and Pressure Vessel Code, Section Xl, Division 1, Subsections IWE and IWL, as modified by NRC final rulemaking to 10 CFR 50.55a published in the Federal Register on October 1, 2004. This Program Plan was developed using the guidance in EPRI's Containment Inspection Program Guide, (ASME Section XI, Subsections IWE and IWL, GC-110698).

The components subject to ASME Section XI, Subsection IWE and IWL requirements are those that make up the containment structure, its leak tight barrier (including integral attachments) and those that contribute to its structural integrity.

Specifically included are Class MC pressure retaining components and their integral attachments, (including metallic shell and penetration liners of Class CC pressure retaining components and their integral attachments), per IWE-1 100; and Class CC reinforced concrete containments and post-tensioning systems, per IWL-1100.

The terms "Class MC" and "Class CC" are used in Section XI to identify components which meet the functional definitions in IWE-1 100 and IWL-1 100; these terms should not be equated with components and items that are designed per ASME Section Ill, Class MC and Class CC rules. Typically, the scope of components and items subject to ASME Section III rules for Class MC Containment vessels and Class CC pre-stressed and reinforced concrete containments extends beyond that of ASME Section Xl, Subsections IWE and IWL.

This Program Plan is effective from January 1, 2007 to December 31, 2016 for Subsection IWE and Subsection IWL activities. IWE and IWL inspections will be Page 1 ofl14

CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT performed according to the schedules shown on Tables 4.1.2.4-1 and 4.2.2.4-1 in ISE-4.

This report includes Period 3 IWE and IWL Inspections performed during the Planned Outage for RF-22 during the Fall of 2015.

The scope of inspections is in accordance with the ISE-4 RF-22 Outage Plan for Interval 2, Period 3 and included the following component inspections:

Entire Accessible Steel Cylinder and Dome Liner

Moisture Barrier at 412' Basement slab perimeter concrete to steel liner. (Augmented)

Concrete Tendon Access Gallery (General and Augmented)

Valve Chambers and Guard Pipes for "A" and "B" RHR and Spray (Augmented)

Penetrations and Hatches

Containment Exterior Concrete Surfaces

Prestress Tendon and Corrosion Protection Systems Design Guideline ST-07, Containment Inservice Inspection Evaluation Criteria, was developed to support this program, and provides inspection criteria used to identify degradation mechanisms requiring documentation as "Recordable Indications". Also included are descriptions of suspect conditions which require evaluation and resolution by the Responsible Engineer.

3.0 INSPECTION PERSONNEL Detailed inspections were conducted by or under the supervision of Quality Control (QC) Lead Personnel and the Responsible Engineer. Each inspector has been qualified to meet the requirements of the VCSNS CISI Program Plan ISE-4 or an equivalent vendor program.

The IWL prestress system tendon inspection and tests as well as the containment concrete examinations were performed by vendor (Precision Surveillance Corporation, East Chicago, IN) qualified inspection personnel using the vendors program and procedures that were reviewed and approved by the Owner, SCEG, as meeting the V. C. Summer Station program requirements.

Precision Surveillance Corporation has been performing the IWL examinations and tests for many years for the Virgil C. Summer Station containment.

The IWE/IWL Responsible Engineer or his representative participated interactively with the QC supervision and inspection personnel. The Responsible Engineer, Dale D. Krause, has a BS Degree in Civil Engineering from Lehigh PAGE 2 of 14

CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT University, with over 15 years of experience in the design, modification, and inspection of Virgil C. Summer Nuclear Station and over 30 years experience in structural engineering in the field of nuclear power plants including nuclear containment design. The Responsible Engineer is a Registered Professional Engineer in the State of Pennsylvania (PE-020392-E) and is the Principal Civil Engineer at Virgil C. Summer Nuclear Station.

4.0 IWE I IWL INSPECTIONS The 2015 IWE and IWL inspections were conducted as an ongoing assessment of the condition of the containment structure. The IWE/IWL inspections and tests were started in October and completed in November 2015.

The IWE inspections included a combination of General Visual, VT-I, and VT-3 examinations performed by qualified QC inspectors. Inspections were performed for all accessible containment liner surface areas using direct line of sight from permanent vantage points. The QC inspector walk downs were used for all containment liner surfaces (including penetrations) as well as the containment liner surfaces in the dome above the spring line. QC inspections were also made on the containment isolation valve containers and guard pipes for the RHR and Reactor Building Spray Systems which are defined as extensions of containment and are located in the Auxiliary Building at Elevation 397'.

It is noted that the exterior surfaces of the carbon steel guard pipes within the penetration sleeves were prepared and recoated during RE 20. The surface of the guard pipes was visually inspected by boroscope prior to and after recoating, during RF 21, and also during current RE 22 inspection in 2015 as designated Augmented Inspections.

A minimum of two coats were applied following the hydrolazing surface preparation. In addition to the required visual Augmented Inspections of the RHR and Containment Spray Penetrations, the augmented inspection of the recoated guard pipe surfaces during RE 20 included visual examination by boroscope of the recoated guard pipes within the penetration sleeves. These areas were reinspected by boroscope during RE 21 and during RE 22.

5.0 INSPECTION PHILOSOPHY The 2015 inspection is an ongoing inspection and assessment program in compliance with ASME Section Xl Subsections IWE and IWL. Previous examinations had identified areas for augmented examination. The augmented examinations were conducted to determine whether continued degradation had PAGE 3 of 14

CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT occurred, the extent of the continued degradation, or if the degradation had stabilized relative to the results of the previous inspection.

6.0 RESPONSIBLE ENGINEER EVALUATION Based on the inspections and examinations during RF-22, no degraded conditions have been identified by the Responsible Engineer evaluation which are considered to be abnormal degradation or of structural function significance.

Additionally, no new conditions which exceeded the ST-07 threshold criteria (i.e.

are likely to experience accelerated degradation or aging) were identified during the inspections.

6.1 IWE Evaluation For the IWE inspections, accessible areas are defined as visible using direct line of sight from permanent vantage points. On the inside of containment there were containment liner surface areas which were previously determined to be inaccessible either due to high radiation or obscured by direct line of sight from permanent vantage points. These inaccessible areas are a small fraction of the total surface area inspected and are also subject to the same environmental and/or service conditions as the much larger representative areas that were inspected.

6.1.1 IWE Augmented Inspections Augmented inspections of the following components have been conducted since damage/degradation was identified during the baseline inspection done in 2000 and were performed again during RF-22.

- Moisture Barrier Integrity

- RHR and RB Spray Penetration Guard Pipes 6.1.1.1 Moisture Barrier Integrity:

All accessible areas of the Containment Moisture Barrier between the perimeter of the basement floor slab at Elevation 412' and the Containment liner were examined by VT-I during RF 22. The Containment Moisture Barrier seals the small gap between the perimeter of the concrete floor slab and the containment steel liner plate. This sealant joint has been subject to inspections and maintenance over the years because it has been observed that minor PAGE 4 of 14

CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT degradation has occurred at the sealant detail along with some light rusting in a few local areas of the RB liner plate typically where the sealant loses adhesion to the liner. The Moisture Barrier seal has been the subject of NRC Information Notice 2004-09 Corrosion of Steel Containment and Containment Liner because typical minor degradation in the sealant and minor rusting on the adjacent liner has been identified at a number of plants. None of the inspection findings documented below reduced the design basis thickness of the RB liner plate nor reduced the capability of the liner to perform the required design basis containment function under the required loads and conditions. The examination was in accordance with the ISE-4 plan and specifically the ASME Code for IWE, Table 2500-1 in the 2001 Edition with the 2003 Addenda.

The Augmented Inspection of the moisture barrier seal performed during RF-22 (Ref. STTS 1412347-001) identified a number of locations where the sealant detail had discontinuities. The locations and lengths of moisture barrier seal requiring maintenance rework were identified in the CR 15-04864. In general the moisture barrier sealant was performing its function as required and the local issues identified during the inspection were typical of local conditions found during each inspection which require some rework to maintain the moisture barrier seal to ensure continued performance to meet the design function.

Sixteen locations were identified ranging in length from about 0.5 feet to 12 feet and are typical of the type of degradation seen during previous inspections which are typically reworked during the RF when they are observed. The longer lengths identified such as one 6 foot length and 12 foot length, were not deficient along the entire length but had several local indications along the length such that the most effective rework was to rework the entire length. The locations were reworked under Work Order 1510797 to the acceptable condition as described in CR 15-04864 and then reinspected.

It is noted that CER 04-1517,CR 08-01993, CR 09-04879, CR 12-05160, and CR 14-02363 documented similar observations found during previous IWE Inspections of the moisture barrier seal and the actions that were taken to perform the maintenance to rework the identified locations to meet design requirements.

Augmented Inspections will continue during each refueling outage to ensure the moisture barrier seal condition is maintained so that it serves its design function to protect the liner from corrosion.

6.1.1.2 RHR & Spray Guard Pipes During the previous RF-22 outage, the Augmented Inspection of the RHR and RB Spray Guard Pipes was performed by Quality Control and Design Engineering. The purpose of the inspection was to continue to monitor issues PAGE 5 of 14

CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND JWL RESPONSIBLE ENGINEER EVALUATION REPORT previously identified during the initial IWE / IWL inspections (Fall 2000). The interspace between the penetration Guard Pipes and the surrounding sleeve pipe has exhibited groundwater in leakage with some corrosion on the exterior of the carbon steel Guard pipes. It was confirmed that the guard pipe thickness met the minimum required design thickness. The completion of the implementation of ECR-50560, "Dewatering System Design and Installation" since the Augmented Inspection during RF 17 had reduced to some extent the groundwater in leakage at the guard pipes which facilitated the rework to prepare and recoat the guard pipes during RF18.

The inspections performed during RF 19 including direct visual and remote visual by boroscope camera found that groundwater in leakage was still occurring resulting in conditions that were found to be supporting corrosion of the guard pipes at some of the newly coated areas from RF 18. This condition was evaluated under CR-I11-03206. The evaluation determined that the guard pipe wall thickness measurements from ultrasonic examination after the hydrolazing prior to recoating during RF 18 indicated thickness with adequate margin to meet design requirements until the next Augmented Inspection during RF 20 in consideration of the rate of corrosion that has been observed over a period of many years on the guard pipes. The proposed corrective actions from CR 11-03206 include performing maintenance work to prepare and recoat the pipe during RF 20. The QC boroscopic inspection of the guard pipes during the initial part of RF 20 identified that the guard pipes recoating to be done during RF 20 had not been completed. The condition was identified and evaluated by CR 12-05536. Subsequently, during RF 20 the guard pipes were prepared, recoated, and reinspected with a boroscope with acceptable conditions observed.

During RF 21 the guard pipes were again inspected as Augmented Inspections.

The condition of the guard pipes was essentially the same as previously left after RF 20 recoating. Some groundwater in leakage was still observed with indications of ongoing corrosion. The condition was reviewed and considered to be acceptable until the next Augmented Inspection during RF 22.

There were no observable changes to the condition of the containment isolation valve containers compared to the previous inspection.

The following historical information is noted. Subsequent, to the early RF-18 Augmented Inspection of the guard pipes, the non-conformance CR-06-03337 Actions 7 and 9 were implemented during RF-18 to perform coating maintenance rework on the exterior surface of the guard pipes. The surface of the guard pipes within the penetration sleeves was prepared by removing existing coatings and corrosion. CR-06-03337 Actions 10 performed inspection to ensure the pipe thickness both locally and in general met the acceptance criteria established in the design basis calculations. The inspection consisted of 100% visual PAGE 6 of 14

CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RlESPONSIBLE ENGINEER EVALUATION REPORT examination of the surface using a boroscope. Local spots of pitting were identified and the thickness determined using measurements taken with a mechanical pit gauge. In addition accessible and representative locations on each guard pipe were identified and thickness measured by Ultrasonic Testing Method (UT) at representative general locations. The Responsible Engineers Technical Work Record attached to CR06-03337 Action 10, documents the UT measurements of wall thickness for the guard pipes and compares the measurements to the minimum required thickness. In all cases there is a good margin between the measured wall thickness and the required wall thickness.

The CR 12-05536 evaluation from RE 20 found the guard pipe thickness to be acceptable and also provides the basis for acceptance of the RF 21 inspection.

RE 22 IWE examination plan included Augmented Examination including 100%

remote visual examination by boroscope of the two RHR and two RB Spray penetration guard pipes. The inspections identified conditions similar to those observed during the RF 21 Augmented Examination, 18 months previously. In leakage of ground water continues via the seismic rattle space separation joint between the exterior face of the Auxiliary Building wall and Reactor Building foundation mat onto the exterior surface of the guard pipes which pass across this separation joint. The in leakage runs downward into the Auxiliary Building along the bottom surface of the guard pipe or along the embedded sleeve pipe which surrounds the guard pipe. The amount of in leakage appears to vary somewhat. Two of the penetrations were relatively dry at the time of inspection but with staining indicating that in leakage does still occur at times. The conditions have resulted in corrosion which continues particularly at the exterior guard pipe surface at the rattle space. The corrosion also follows along the bottom of the guard pipe exterior surface outward into the Auxiliary Building where the groundwater runs along the guard pipe into the building during cycles of wetting and drying. CR 15-05594 identified and evaluated the observed conditions. Ultrasonic Testing was performed at representative guard pipe corrosion areas to confirm that the minimum required guard pipe wall thickness was available and would be remain available taking into account the observed rate of corrosion until the next Augmented examination during RF 23 in 18 months. The minimum available wall thickness determined from the Ultrasonic Testing was found to have at least a margin of 35% greater than the minimum required design basis thickness. The corroded areas will be reworked again by suitable surface preparation and stabilization followed by the reapplication of a suitable coating system. Previous attempts to eliminate groundwater in leakage by lowering the groundwater via dewatering have not been entirely successful.

Therefore, the plan is to rework/recoat the guard pipe exterior surfaces to the extent feasible to minimize progress of the corrosion whije keeping the guard pipes as Augmented Examinations to be done each Refueling outage on an 18 month cycle. Summarizing, the guard pipes were found to be acceptable with minimum measured thickness exceeding the minimum required wall design PAGE 7 of 14

CONTAINMENT INSERVICE JINSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT, thickness by a reasonable margin and will remain acceptable until the next inspection in 18 months.

6.1.2 IWE Other Inspection Results There were no other IWE inspection findings that indicated actual or potential degradation of the components within the IWE inspection boundary scope.

6.1.3 IWE Reportable Conditions There are no IWE Reportable Conditions as a result of the RF-22 year 2015 inspections.

6.1.4 IWE Reportable Conditions Requiring Augmented Inspections None of the results of the IWE inspection were found to exceed the evaluation criteria of Design Guide ST-07 or determined to be of concern that could potentially progress to an unacceptable structural condition prior to the next regularly scheduled surveillance in approximately 18 months during RF-23 (Spring 2017).

6.2 IWL Evaluation 6.2.1 IWL Augmented Inspections Augmented Inspection of the Containment Tendon Access Gallery (TAG) has been conducted during each inspection following its identification for Augmented Inspection during the baseline inspection in 2000. The inspection included previously identified issues from the IWE / IWL inspections (Fall 2000) and subsequent Augmented Inspections (2002, 2003, 2005, 2008, 2009, 2011, 2012, and 2014).

The following were examined:

-Corrosion build-up and leaching on the outer TAG wall near Tendon V-I5 (Ref. STTS 1412607-002)

-Concrete leaching at several locations within the TAG (Ref. STTS 141 2607-001)

PAGE 8 of 14

CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT The corrosion build-up was inspected by VT-I examination and determined to be comparable with the results of the last inspection during RF 21 in 2014 with little or no change from the previous inspections.

The entire TAG was inspected by General Visual examination during RF 22 for housekeeping and for any changes in the amount of concrete leaching. The amount of groundwater in leakage remained minimal and additional accumulation of leaching materials was determined to be about the same as observed during previous inspections. Sumps remained clear of debris to allow for drainage if required.

The General Visual examination of the Tendon Access Gallery performed during RF 22 by Quality Control and Design Engineering identified several areas of leaching and discoloration but with none or minor increase from the previous inspection during RF 21. Therefore, the condition of leaching that has been evaluated previously is unchanged and acceptable. Only minor traces of grease seepage were observed coming across the gasket at a few vertical tendon end caps. The quantity was insignificant and did not represent any degradation of the tendon corrosion protection system.

Summarizing, the Tendon Access Gallery area housekeeping was found to be acceptable. The corrosion near tendon V-i5 on the outer wall will remain on the list for Augmented Inspection by VT-I. The overall General Visual Examination of the entire north and south parts of the Tendon Access Gallery area shall continue to be performed during each refueling outage. This General Visual Examination is important not only to confirm conditions of the reinforced concrete, but also to check that there is no excessive unexpected leaking of the corrosion protection grease from any of the vertical tendon end caps and gaskets. The General Visual examinations of the TAG for concrete leaching and general housekeeping will continue as General Visual examinations under the ASME Section XI IWE/IWL program to be performed as part of the Augmented examinations every Refueling Outage.

6.2.2 IWL Prestress Tendon System Evaluation The Ninth Period Tendon Surveillance (3 5 th year) was performed immediately prior to and during RF 22 between October and November 2015. The surveillance work activities were performed by Precision Surveillance Corporation (PSC) to approved Vendor Procedures that met the requirements of the V.G. Summer tendon inspection program documents including Engineering Services Specification SP-228," Surveillance of Reactor Building Post Tension System" and Surveillance Test Procedure STP-160.001, "Containment Tendon Test."

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4 CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT The complete details of the inspection including procedures, results, and conclusions are contained in the PSC Document REP-I1120-510 Rev. 0 "Final Report for the 3 5 th Year (9th Period) Tendon Surveillance at V.C. Summer Nuclear Station." The report is on file with Surveillance Test Task Sheet STTS-01412605. The following is a summary of the inspection results.

Three tendons from each tendon group (vertical, dome, and hoop) were randomly selected for testing. The predicted liftoff forces were calculated for each tendon in accordance with the methodology of Regulatory Guide 1.35.1. The measured liftoff forces for 8 of the 9 surveillance tendons were found to be above 95% of the predicted lift off force values as required by IWL-13221 .1(a). Dome tendon D216 average liftoff force was 1099 kips or 93.7% of the average predicted liftoff force. The condition was documented by CR-I15-04979. In accordance with the approved procedure, liftoff forces were taken for the adjacent tendons D215 and D217 with both tendons found to have greater than 95% average predicted liftoff force. D216 was the only tendon in the 9 tendon surveillance population with less than 95% average predicted liftoff force.

Therefore, the liftoff force for D216 was acceptable.

The group average liftoff force for all as-found liftoff values as well as the individual liftoff values were all above the required average minimum design forces for the respective tendon group and are acceptable.

GROUP GROUP AVERAGE (KIPS) MINIM/UM DESIGN AVERAGE (KIPS)

Dome 1085.5 1063 Hoop 1067.1 1000 Vertical 1284.7 1160 The results of the regression analyses for each of the three tendon groups confirm that the minimum group average prestress will be available for each group through the period until the next tendon surveillance for year 40. The regression (trend) analysis results using the current as well as previous liftoff data show all groups forces staying above the required minimum design force beyond the next regularly scheduled surveillance (4 0t year).

Group Forecast Years after Initial Prestress When Group Mean Will Fall

_________Below Design Minimum Average Group Prestress Dome 45 Hoop > 60 Vertical > 60 It is currently projected based on these results that the Dome tendon group will be retensioned to a level that ensures the minimum group average prestress is PAGE 10 of 14

CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT.

available beyond the plant license (currently 60 years) at approximately the next scheduled tendon surveillance in year 40.

All of the other required surveillance and test criteria for the tendons including tendon force and elongation, wire samples inspection and testing, grease samples testing for water and impurities, tendon anchorage hardware and adjacent concrete areas, and corrosion protection grease were acceptable.

The following were identified and evaluated to be acceptable conditions not indicative of abnormal degradation nor needing additional augmented inspections:

The horizontal tendon H46CB selected for detensioning was found upon removal of grease cap and cleaning of grease from anchorage with one protruding wire from the shop end and one from the field end. In both cases the projection was only by a small amount. Continuity testing was performed on each protruding wire and determined the wires were fully effective. CR1 5-05209 evaluated the condition. Upon retensioning H46CB, one additional wire, not previously of the two identified upon detensioning, was observed to be protruding by a small amount. This wire had also been confirmed to be intact by continuity testing.

Tendon H46CB upon detensioning had one sample wire removed for testing.

Therefore, tendon H46CB was left with 169 of 170 effective wires. No additional protruding/missing button heads!/or missing wires were detected on any of the other surveillance tendon population. The condition was evaluated and concluded to be acceptable by CR 15-05387.

Individual sample wire tensile testing from the one wire removed from each of the three detensioned tendons (one from each group) found that the strength for all wire sample tested exceeded the minimum specified ultimate strength 240 ksi and the corresponding wire elongation at failure exceeded the 4% minimum elongation required by the ASTM material specification for the high strength wire.

Augmented scope of the surveillance included performing grease leak repair by replacing gaskets for two dome tendon ends and three vertical tendon lower ends which were identified by the inspections to be inactive leakage with a minor amount grease showing across the end cap gasket.

Summarizing, the surveillance and tests confirmed that the prestress tendon system and its corrosion protection system are performing the license basis design functions without indication of abnormal degradation and will continue to perform those functions through the next regularly schedule surveillance at the 40th year.

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CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT 6.2.3 IWL Containment Exterior Concrete Surfaces The general visual inspection was performed by qualified Precision Surveillance Corp. inspection personnel concurrent with the tendon system surveillance during RF 22. The scope included all accessible exterior concrete surfaces of the Reactor Building both above and below roof lines, the dome, and the ring girder.

The inspections were performed in accordance with approved PSC procedures that met the requirements of the Owners program. The results of the inspection are documented in the PSC Document REP-1120-510, Revision 0, "Final Report for the 3 5 th Year Tendon Surveillance at V.C. Summer Nuclear Station." The Inspection findings were minor and were assessed as minimal or no change observed from the previous surveillance performed concurrent with the 30t year tendon surveillance. None of the inspection findings represented abnormal degradation of the concrete surface. Summarizing the concrete inspection for the containment exterior met the acceptance criteria and found the overall concrete in good condition demonstrating continued durability.

The concrete inspection was performed in accordance with STP-803.006D-RB under work order 1013932-001.

6.2.4 IWL Reportable Conditions No reportable items or items indicative of abnormal degradation were identified by the IWL inspections for:

(1) Augmented Inspection of the Tendon Access Gallery.

(2) 35 th Year surveillance of Prestress Tendon System.

(3) General Visual examination of all exterior concrete surfaces of the containment structure performed concurrent with the Prestress Tendon System surveillance.

6.2.6 IWL Reportable Conditions Requiring Augmented Inspections None.

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CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT 7.0

SUMMARY

AND CONCLUSIONS The ASME Section Xl IWE inspections performed for VCSNS during period October-November 2015 concurrent with RF 22 are the continuation of the ongoing required inspection of the containment structure. These inspections provide the necessary basis for comparison with future surveillance results.

All recordable indications identified as meeting or potentially exceeding acceptance criteria were evaluated by the Responsible Engineer and found to be acceptable with no impact on the capability of the Reactor Containment Building to meet its required license basis design functions.

Minor areas and evidence of groundwater inflow leakage and some concrete leaching formation that have been observed during previous inspections were observed during the General Visual examination of the Tendon Access Gallery.

These conditions were unchanged compared to the previous examination during RE 21. None of these observations impact the design function of the reinforced concrete containment structures. The Tendon Access Gallery will continue to be reinspected by General Visual Examination and also by Augmented Inspection using VT-I for the local corrosion buildup previously identified in 2000 on the outer TAG wall near tendon V-I15.

The Augmented Inspection performed during RF 22 found the Moisture Barrier seal required some rework to restore it to required condition in 16 locations around the perimeter of the containment basement floor. The conditions were evaluated and reworked under CR-I15-04864. The observations and reworked conditions were typical and comparable to previous inspection observations identified by CRs in the past. The observations included local spots of debonding of the sealant from the steel liner and concrete floor.

The following areas shall remain on the Augmented Inspection list:

eIWE - Moisture Barrier Integrity o IWE - Guard Pipe Containment Boundary at Auxiliary Building Elevation 397,

IWL - Tendon Access Gallery Corrosion Each of the Augmented Inspection areas shall be inspected during each refueling outage (at approximately 18 month intervals) to ensure that any structural degradation should it occur during the period between inspections will be examined and evaluated.

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A CONTAINMENT INSERVICE INSPECTION - 2015 ASME SECTION XI, SUBSECTIONS IWE AND IWL RESPONSIBLE ENGINEER EVALUATION REPORT The IWL inspections performed during RF 22 in Fall 2015 included:

oIWL- Prestress tendon system.

o IWL - Reactor Building exterior concrete surfaces.

These IWL inspections performed at a 5 year inspection interval found that the containment structure is performing its design function as required and experiencing no abnormal degradation.

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~1 George A. Lippard Vice President, Nuclear Operations 803.345.4810 A SCANA COMPANY March 3, 2016 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

Dear Sir / Madam:

Subect: VIRGIL C. SUMMER NUCLEAR STATION (VCSNS), UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 INSERVICE INSPECTION (181) OWNER'S ACTIVITY REPORT (OAR) FOR REFUELING OUTAGE 22 Attached is the Inservice Inspection (1S1) Owner's Activity Report (OAR) covering ISI activities associated with the second outage for the first period of the fourth ISI interval and the second outage for the third period of the second Containment Inservice Inspection (CISI) interval. This report is submitted pursuant to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (2007 Edition through 2008 Addenda), Section Xl, Rules for Inservice Inspection of Nuclear Power Plant Components, Article IWA-6000, Records and Reports, and ASME Code Case N-532-4, Repair/Replacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission, Section Xl, Division 1.

Should you have any questions, please call Mr. Bruce Thompson at (803) 931-5042.

WLT/GALlwm Attachments:

1. Inservice Inspection Owner's Activity Report for Refuel 22, Report Number 21

2. Containment Inservice Inspection -2015 ASME Section Xl, Subsections IWE and IWL Responsible Engineer Report c: K. B. Marsh S. A. Williams S. A. Byrne NRC Resident Inspector J. B. Archie K. M. Sutton N. S. Camns NSRC J. H. Hamilton RTS (LTD 323)

J. W. Williams File (810.19-3, RR5300)

W. M. Cherry PRSF (RC-16-0034)

C. Haney V.C.Summer Nuclear Station

P.O.Box 88.* Jenkinsville, SC. 29065 ;,F (803) 941-9776 Aoqq