ML13289A183

Response to Request for Additional Information Concerning Relief Request No. RR-III-10

ML13289A183
Person / Time
Site:	Summer
Issue date:	10/11/2013
From:	Gatlin T South Carolina Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MF1848
Download: ML13289A183 (22)

Text

Thomas D. Gatlin Vice President,Nuclear Operations 803.345.4342 October 11, 2013 A SCANA COMPANY U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS), UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONCERNING RELIEF REQUEST NO. RR-II-10 (TAC NO. MF1848)

Reference:

1. Thomas D.: Gatlin, SCE&G Letter to Document Control Desk, Request Relief from ASME Code Requirements in VCSNS 3 fd Ten Year Inservice Inspection Interval, dated May 6, 2013 [ML13129A178]

2. Robert E. Martin, NRC Letter to Thomas D. Gatlin, SCE&G, Request for Additional Information Concerning Relief Request No. RR-111-1O (TAC No.

MF1848), dated September 11,2013 [ML13252A177]

South Carolina-Electric & Gas Company (SCE&G) received a NRC letter dated September 11, 2013 (Reference 2), requesting additional information (RAI) regarding the Virgil C. Summer Nuclear Station Unit 1 relief request from ASME code requirernrnts in the VCSNS 3 rd Ten Year Inservice Inspection Interval (Reference 1). SCE&G has reviewed the request for additional information and hereby submits the attached response.

If you have any questions or require additional information, please contact Mr. Bruce Thompson at (803) 931-5042.

Very truly yours, Thomas D. Gatlin TS/TDG/wm Enclosure - Response to RAI Concerning RR No. RR-III-10 (TAC No. MF1 848)

Attachments (2) c: K. B. Marsh S. A. Williams S. A. Byrne NRC Resident Inspector J. B. Archie K. M. Sutton N. S. Carns NSRC J. H. Hamilton RTS (CR-12-05348)

J. W. Williams File (810.19-2)

W. M. Cherry PRSF (RC-13-0157)

V. M. McCree Ac~q7 Virgil C.Summer Station . Post Office Box 88 - Jenkinsville, SC - 29065 . F (803) 941-9776

Document Control Desk Enclosure CR-12-05348 RC-13-0157 Page 1 of 6 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) Unit I DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ENCLOSURE VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. I - RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONCERNING RELIEF REQUEST NO. RR-111-10 (TAC NO. MF1848)

Document Control Desk Enclosure CR-12-05348 RC-1 3-0157 Page 2 of 6 REQUEST FOR ADDITIONAL INFORMATION REGARDING VIRGIL C. SUMMER NUCLEAR STATION UNIT 1.

SOUTH CAROLINA ELECTRIC & GAS CO.

RELIEF REQUEST NO. RR-111-10 DOCKET NO. 50-395 (TAC NO. MF1848)

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the information provided by South Carolina Electric & Gas Co. (SCE&G, the licensee) for Virgil C.

Summer Nuclear Station, Unit 1 (VCSNS) Relief Request No. RR-1II-10 in its letter dated May 6, 2013 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML13129A178), and has determined that additional information is necessary to complete the review of Relief Request No. RR-III-10.

Based on the staffs review of Relief Request RR-III-10, please provide responses which address the following requests:

RAI 1

The NRC was not able to locate a prior submittal from SCE&G requesting relief from essentially 100% inspection coverage for the VCSNS pressurizer surge nozzle.

Given that the current design configuration for the VCSNS surge nozzle has been present since the original plant design, and given that VCSNS is currently in the third ISI interval, provide a copy of the previous Relief Request for inspection of the surge nozzle. If inspection relief for essentially 100% coverage was not previously requested, explain why.

SCE&G RESPONSE The Preservice Inspection Program contains a statement that a relief request should be submitted due to examination being limited by geometric configuration to volumetric examination from the shell side. It appears that VCSNS did not request relief for limited coverage during the 1st or 2nd Intervals. The reason for this is unknown; however, it is suspected that the prior ISI program owners may have interpreted the statement, "to the extent practical within the limitations of design, geometry...," to have meant no relief was required. This, according to Information Notice (IN) 98-42, "Implementation of 10 CFR 50.55a(g) Inservice Inspection Requirements," was a fairly common issue prior to release of the IN as it warranted the creation and release of generic communications.

Document Control Desk Enclosure CR-12-05348 RC-13-0157 Page 3 of 6

RAI 2

The Wesdyne Ultrasonic Examination Reports included in Attachment 2 of the Relief Request identify that no recordable indications were present in any of the scans for the surge nozzle. Have any indications previously been detected on the pressurize surge nozzle-to-vessel weld that were evaluated to be either relevant or non-relevant indications? If any indications were detected, discuss the inspection during which the indications were found and the disposition of the indications.

SCE&G RESPONSE No indications were noted on any of the 3 prior exams (PSI, ISI Interval 1, and ISI Interval 2). Coverage maps for ISI in intervals 1 and 2 match the coverage for interval 3.

RAI 3

Has a visual examination been performed on the pressurizer surge nozzle-to-vessel weld? If so, what were the results of the examination?

SCE&G RESPONSE:

The Pressurizer lower head and heater sleeves are visually inspected (VT-2) with insulation on as part of both the PTP-151.001A Inspection for Boric Acid Corrosion and the STP-250.001A Reactor Coolant System Leakage Test each refueling outage while the Reactor Coolant System is at pressure. Any indication of boron and/or leakage on the lower pressurizer head insulation requires a bare metal visual inspection under the insulation to be performed per procedure. The exam results have been satisfactory for the pressurizer lower shell area above the surge nozzle-to-shell weld with no indications of leakage in any part of the lower pressurizer area; see Attachment II of this submittal. (PTP-151.001A results are pages 2 through 8 of Attachment II and STP-250.001A results are pages 9 through 13 of Attachment II.)

RAI 4

Provide an inspection scan coverage map that identifies the areas of the examination volume that were missed.

SCE&G RESPONSE:

The inspection scan coverage map identifying the areas of the examination volume that were missed was provided with the initial relief request (Reference 1 listed in cover letter). Attachment I of this submittal provides a sketch of the inspection scan coverage map with each direction's volume achieved inside of the respective dotted lines. The

Document Control Desk Enclosure CR-12-05348 RC-1 3-0157 Page 4 of 6 coverage in the axial scanning directions, AxHd = Axial Head (scanning from the head towards the nozzle) and AxNz = Axial Nozzle (scanning from the nozzle towards the head), had limited coverage due to the physical obstructions shown on the exam sketch.

RAI 5

Identify whether the pressurizer surge nozzle receives a visual (VT-2) examination in conjunction with the Class I System Leakage Test conducted during each refueling outage in accordance with ASME Section Xl requirements to compliment the limited examination coverage. If so, identify the number of such examinations that took place during the third interval.

SCE&G RESPONSE:

The entire pressurizer is inspected as part of the STP-250.001A Reactor Coolant System Leakage Test each refueling outage. The exam is performed with insulation on when the system is pressurized greater than 350 psig coming out of a refueling outage. Interval 3 outages were Refueling Outage (RF)-15, RF-16, RF-17, RF-18, RF-19, and RF-20.

RAI 6

There are discrepancies between the text in the Relief Request and the data sheets provided in Attachment 2 related to search unit angles for each scan direction as well as coverage achieved. As such, please clarify the following:

a. Attachment 2, pages 1 and 2 indicate that a 55 degree ( 0 ) search unit was also used in the surge nozzle examination, but this search unit was not described in the relief request. Explain why and how this search unit was used in the surge nozzle weld examination.

SCE&G RESPONSE

a. Pages 1 and 2 of Attachment 2 of the relief request are actually the inner radius exam for the pressurizer surge nozzle. The EXAMINATION WELD/AREA is listed as 1-2100A-81R, where the "IR"denotes that this is the inner radius examination of the pressurizer surge nozzle. Pages 1 and 2 were included to demonstrate the completeness of the surge nozzle weld exam since pages 3 through 6 pertained to the nozzle-to-shell weld exam only.

Document Control Desk Enclosure CR-12-05348 RC-13-0157 Page 5 of 6

b. Attachment 2, page 3 indicates that there were no scan limitations for the 00 search unit. However, the Code Coverage Achieved is shown as 75.5%.

Explain this apparent discrepancy.

SCE&G RESPONSE

b. The Code Coverage Achieved box shows the total examination combined code coverage achieved for the weld/area examined. The 0 degree search unit was used for contour mapping the weld ID. The exam coverage shown on page 6 shows that only 2 transducers were used for the code exam: 45 degrees and 60 degrees.

c. Attachment 2, page 4 indicates that the Code Coverage Achieved for the 450 search unit was 75.5%. However, Attachment 2, page 6 and the text of the relief request indicates the coverage was 51% for this search unit. Explain this apparent discrepancy. Also, why is the scan area for this search unit indicated as both parallel and perpendicular to the weld?

SCE&G RESPONSE

c. The Code Coverage Achieved box shows the total examination combined code coverage achieved for the weld/area examined. The ultrasonic (UT) exam sketch sheet, page 6, states: "weld coverage based on % of weld material scanned from both directions with both transducers perpendicular and parallel to the weld (4 directions)."

This means that each percentage of coverage is based on the amount of material scanned by the combination of the 45 degree and 60 degree transducers coverage areas. In the clockwise (CW) and counterclockwise (CCW) directions (parallel to the weld centerline), the examiners were able to get 100 percent coverage with the combination of the transducers. In the Axial directions perpendicular to the weld centerline (Ax Hd - from the head towards the nozzle, and Ax Nz - from the nozzle towards the head), the examiners were only able to get 51 percent of the required examination volume in each direction with the combined transducer coverage due to the physical obstructions shown. In the comments section on page 6, it can be seen how the total combined coverage is calculated. The code requires scanning in two directions when looking for axial indications, and two directions when looking for circumferential indications, and with multiple angles.

d. Attachment 2, page 5 indicates that the Code Coverage Achieved for the 600 search unit was 75.5%. However, Attachment 2, page 6 and the text of the relief request indicates the coverage was 51% for this search unit. Explain this apparent discrepancy. Also, why is the scan area for this search unit indicated as both parallel and perpendicular to the weld?

Document Control Desk Enclosure CR-12-05348 RC-1 3-0157 Page 6 of 6 SCE&G RESPONSE

d. The Code Coverage Achieved box shows the total examination combined code coverage achieved for the weld/area examined. The UT exam sketch sheet, page 6, states: "weld coverage based on % of weld material scanned from both directions with both transducers perpendicular and parallel to the weld (4 directions)." This means that each percentage of coverage is based on the amount of material scanned by the combination of the 45 degree and 60 degree transducers coverage areas. In the clockwise (CW) and counterclockwise (CCW) directions (parallel to the weld centerline), the examiners were able to get 100 percent coverage with the combination of the transducers. In the Axial directions perpendicular to the weld centerline (Ax Hd - from the head towards the nozzle, and Ax Nz - from the nozzle towards the head), the examiners were only able to get 51 percent of the required examination volume in each direction with the combined transducer coverage due to the physical obstructions shown. In the comments section on page 6, it can be seen how the total combined coverage is calculated. The code requires scanning in two directions when looking for axial indications, and two directions when looking for circumferential indications, and with multiple angles.

Document Control Desk Attachment I CR-1 2-05348 RC-1 3-0157 Page 1 of 2 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT I DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT I COVERAGE AREA OF EXAMINATION VOLUME THAT WAS MISSED

I I UUU

%Wel I Compoent iD Number. 1-2100A-8.

~dwPum*n

~, R~~g 51% 51%

AxHd AxNz Wdd covwg bood an %ofwdd mwuu scw)ad fum bc~&dkmn wft bo* tmank pepe"ui a loud nl lofw o wed (4d dmsln)

Document Control Desk Attachment II CR-12-05348 RC-13-0157 Page 1 of 13 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT II VISUAL EXAMINATION OF PRESSURIZER SURGE NOZZLE-TO-VESSEL WELD RESULTS

PTP-151.001A Attachment I Page 1 of 1 Revision I PMTS # 2 -9 Aranspection.Repor Step Area/Component Sat Unsat CER #

tet() (') (if required) 6.1.1 Reactor Vessel Head area (including head ventilation ductwork) 6.1.2 Reactor Coolant Pumps V 6.1.3 Steam Generator Primary Side Manways 6.1.4 Pressurizer Spray Valves /

6.1.5 Pressurizer Safety Valves /

6.1.6 Pressurizer Manways /

6.1.7 All accessible Reactor Coolant Piping 6.1.8 All Elevations - General Area * . / '

EfIS 5.4.1 The Precautions listed in Section 2.0 have been reviewed with all per he performan of the test.

SSlA4.ignature be is'e 5.4.2 been satisfied.

Page 2 of 13

PTP-151.001A ORIGINAL Attachment I Page 1 of I Revision 1 PMTSý7/ 8PIPYOO Area Inspection Report CER Step Area/Component Step() Unsat Sat () (if required) 6.1.1 Reactor Vessel Head area (including head ventilation ductwork) 6.1.2 Reactor Coolant Pumps 6.1.3 Steam Generator Primary Side Manways 6.1.4 Pressurizer Spray Valves 6.1.5 Pressurizer Safety Valves 6.1.6 Pressurizer Manways 6.1.7 All accessible Reactor Coolant Piping 6.1.8 All Elevations - General Area 5.4.1 The Precaution, listed in Section 2.0 have been reviewed with all person i o in the performapce of. the test.

/4I W 5.4.2 have been satisfied.

Page 3 of 13

A I AQ ,SP-216 I' ATTACHMENT III PAGE 1 OF I REVISION I INSPECTION REPORT FOR BORIC ACID LEAKAGE Visual examination for Boric acid leakage Equipment ID 2't2-/V(/z. Af#)4,JSf) System. RjC Location ,Z[3 ,26 Code Class: I (IWB)L_

Drawing /O-- 2 (IWC) 3 (IWD)_

Description of Leakage , , &4A,. V',5.6,a-,d, Method of examination: Direct _ Remote Video Insulated Non-insulated Insulation removed Cracks: Accepted Rejected NA ."

Corrosion: Accepted Rejected _NA ,N Bolt Coatings: Accepted Rejected NA7.7 Leaks near bolting: Accepted Rejected NAP,"

Acceptance Criteria: IWA 5250 (b) Pre-Service ln-Service...--*§ Comments Inspected by7( 9A* Z" /7 Date - Level Reviewed by 715/'Y"4 Date ANII AI*' Date Page 4 of 13

f 4

PTP-151.001A Attachment I Page 1 of 1 Revision 2 Ak oe Reactor Buildina Area Insoection Reo we # 10.04-4i2A-001.,

5.4.1 The Precautions listed In Section 2.0 have been reviewed with all personnel involved in the performance of the test. .

b c Ls . /

5.4.2 The required Initial Conditions for this test have RB~c~rIMc BR Sku~tur DaI

" 88/CRSIWCC SRODi nature Date step Sat Unsat CR# QC Inspector StepAreaCompnent(v') (v') UNSAT RoquNredif Initial & Date Reactor Vessel Upper Head Area &-/ . P4 6.1.1 (including head ventilation dudtwork) J O/-/&2/

O- ,

6.1.2 I Reactor Vessel Lower Head, Inoore Pit _ _ _ ____"_

61.3 ...... Reactor Coolant Pump; XPPO030A "

Reactor Coolant Pump; XPPD030B Reactor Coolant Pump; XPP0O300 .. /___ ..

Steam Generator Primary Manway 6.1.4 SIG A INLET Steam Generator Prlmary Manway SIG A OUTLET Steam Generator Primary Manway.

SIG B INLET ._ _

Steam Generator Primary Manway 8 1 B OUTLET Steam Generator Primary Manway SIG C INLET _

Steam Generator Primary Manway SIG C OUTLET 6,.1.5 ' P r'essurlzer Sn a Valve: PCV *44 C .... . . . . . ._._... .

PCV 4440

.6'16. Pressurizer z Valves;r XVSa010Anwa xvsg010B XV8600i0C' Z

Pressurizer §pryVale- PCV444C Pressurizer Relief Valvesa; xveGo000B/

XVe le000A xvG8oooc .

Pressourizer POgV: PCV444B /

PCV446A / .. .

PCV445B 6.;17./ Pressurlizer Manway -

-ll*l .l

. =* f U " -L,**-

& Heater

, Sleeves .

-6.1.8 " Pressur'izer Lowe-rHead 61.9 All accessIble Reactor Coolant PipIng z

.1.10 General Area Inspection of RB 463' and below Remarks:

Page 5 of 13

V.C. Summer Nuclear Station Page I of 2 10/912013 CR-11-01739 CR Print 17:16A40 Overall Section Created 411612011 State Closed Safety Class SR Discovered 4116/2011 System RC REACTOR COOLANT Repetitive N Occurred 4/16/2011 Criticality Crit I Group 9 Identified By MEL BROWNE Primary Equipment Equipment Name XTKO024 PRESSURIZERASSY Additional Equipment Brief Description During the Reactor Building walk-down Inaccordance with PTP-151, the QC inspector observed an unknown white material on a pressurizer heater tube. The inspection was performed from the floor of the RB looking up to the bottom of the pressurizer.

Results reported by Roy Caban Detailed Description During the Reactor Building walk-down In accordance with PTP-151, the QC inspector observed an unknown white material on a pressurizer heater tube. The inspection was performed from the floor of the RB looking up to the bottom of the pressurizer.

Results reported by Roy Caban Immediate Action Reported condilion In accordance with SAP-1 100 Screened by AB155ý17 D-isposltioner PSE-Eval Overall Approval PSE Overall Concurrence NON NO Operability Req? Y Reportability Req? Y Evaluation Req? V MRT Y Eval Type D- OTH CR Category - Category 3 PSRC Revw N Meeting No.

Screening Comments Screening Due 4/1712011 Actual Date 4/17/2011 Mlsc Trend Codes Associated Work Orders 1105050 vent Description ... . .. Cause Code DecrI ptin ause Group .

ft-19 R e N ple .OT APPLICABLE IIII.II Appicable. ......

... I Page of 6 of 13

r-crJp"nt V.C. Summer Nuclear Station Page 2of 2 10[/12013 CR-11-01739 CR Print 17:B1:40 Evaluation Section Due Date 51/712011 Current State Closed Assign To SHANNONALI Comments 10 CRF21 N Concurring Group Route Group PSE-Eval D - Other CAUSE EVALUATION CR-1 1-01739 Evaluation Performed By. Shannon All Note: In cases where an evaluation is not performed, the evaluator should state why not. (ie. Work order generated)

1. Provide a statement of what happened. Provide a sound basis for actions that may be generated During the Reactor Building walk-down in accordance with PTP-151, the QC Inspector observed an unknown white material on a pressurizer heater tube. The inspection was performed from the floor of the RB looking up to the bottom of the pressurizer.

2. Provide a statement of probable cause (s):

a. The probable cause (s) of this event was seal penetrant on the welded areas

3. Propose Corrective Actions: Document proposed actions in the action section or Document below actions already taken to resolve the Issue. The residue found not to be boron. No further actions are required.

Attachment List - Overall CR ID cR-41-01739 Keyword screening Location

• Text Contacted Control Room workd control center. Inspection performed from 412' looking up to bottom of pressurizer.

Impossible to determine whether the white material IsBA residue without further investigation.

Page 7 of 13

PTP-151 .OOIA Attachment I Page 1 of 1 Revision 2 wo# ____o'__ _

Reactor Buildina Area Inspection Report 5.4.1 The Precautions listed in Section 2.0 have been reviewed with all per onnel involved in the performance of the test. /,;<, SS,~RS, 7 RISigatur /WC SS,"CRS, or-WCC SRO Signature DateA 5.4.2 The required Initial Conditions for this test have been satisfied. CHG J6,I Wl-I A

.  ! L L i II " i I ! I# " - I SS, CRS, or WCC SRO Signature Date!

Sat Unsat CR # OC Inspector Step Area/Component (0) () Required if Initial & Date I UNSAT Reactor Vessel Upper Head Area 6.1.1 (including head ventilation ductwork) ._ ..

6.1.2 Reactor Vessel Lower Head, Incore Pit I .- , ie'/s/v.-

6.1.3 Reactor Coolant Pump; XPP0030A Reactor Coolant Pump; XPPO030B. ____._,_ C.(i ,7 I Reactor Coolant Pump; XPP0030C i _-___,.

Steam Generator Primary Manway 6,1.4 SIG A INLET - /o/,.

Steam Generator Primary Manway .

SIG A OUTLET J 0 /

Steam Generator Primary Manway SIGB INLET _ ___,_ .._ i___ ,_-.

Steam Generator Primary Manway -

B/G B OUTLET Steam Generator Primary Manway S/GC INLET 7:

Steam Generator Primary Manway S/G C OUTLET 6.1.5 PressurizerS pjapValve: PCV444C ie to-3-/'...

PCV444D " jo-a"' 1.-

6.1.6 Pressurizer Safety Valves; XVS8010A V to_-1) -_ 2.-.

XVS8O1OB V/o-7'~-

XVSB010C _____ fb -2~-

Pressurizer Rlief Valves; XVG8000A . .. r XVG8000B XVG8000C __" __ - #o--

Pressurizer PORV: PCV444B _._________ ,/ l/

PCV445A PCV4458 7/" /

_/t&-t3-i2-6.1.7 Pressurizer Manway _/___/_,___2-6.1.8 Pressurizer Lower Head & Heater Sleeves i ct -o- 2-6.1.9 All accessible Reactor Coolant Piping ________!___ I_ >______-_____

6.1.10 General Area Inspection of RB 463' and below A s'e-rsv..--l,- ct II -j')e.

b coYr't 3o. .0SP'o-ys- Vf Page 8of 13

STP-250.001A Attachment II Page lof2 Revision 2 STTS#

Reactor Coolant System Leakage Test Data Sheet System

Description:

Drawing Number:

Reactor Coolant System See boundary Enclosure(s) listed below. Date: \ .

Type of System: Insulated Class: 1 Type Test: System Leakage Test 5.4.1 The Precautions listed Section 2.0 have been discussed with all ._

personnel involved in the performance of the test. /

S CGR 5.4.2 The required Initial Conditions for this test have been satisfied.

SS/CRS/TUS Signature Date 6.1.2 O*)to Hold T'ime Start \-

6.1.2 The piping inside the boundaries listed below is at normal operating pressure for 100% rated reactor power.

,vo 1J,,I 106 6.1.3 6'4OHold Time End S1/WSf DATE VT-2 Visual Test Inspected within boundary of enclosure(s). YES [* NO EL Remark No.

Boundary Enclosure(s) _ . .-. b-Total Leakage L (Tech. Spec. 3.4.6.2.bit 10 gpm)

All Section 8.0 Acce~tance Criteria for this test have been satisfactorily met. Yes V' No_

VT-2 Qualified Test Performer Datej,-zDo ,

-DateZLýv-?-,'

Date

,a,~

Remarks: AML - ,oo'tp.. -7, Continuation Sheet Required Yes El No "

Page 9 of 13

STP-250.001A Attachment II

~ Page 1 of 2 WORKING COPY S'ITS R # 4g,#'107-,,D/

2 Reactor Coolant System Leakage Test Data Sheet System

Description:

Drawing Number.

Reactor Coolant System See boundary Enclosure(s) listed below. Date: e. .-0 Type of System: Insulated Class: 1 Type Test: System Leakage Test 5.4.1 The Precautions listed Section 2.0 have been discussed with all personnel involved in the performance of the test. _ /I _-_-_

SS/,GS/TUS rignaJur Date 5.4.2 The required Initial Conditions for this test have been satisfied. *r =--9, SS/CRS/TUS Signature Date 6.1.2 The piping inside the boundaries listed below is at normal operating pressure for 100% rated reactor power.

6.1.2 o4L1 Hold Time Start 6.1.3 c Hold Time End SS(CRS/TUS DATE VT-2 Visual Test Inspected within boundary of enclosure(s). YES I NO - Remark No.

Boundary Enclosure(s) 6 9L_-. - j d c_ & 1 pW - -

Total Leakage , o4? (Tech. Spec. 3.4.6.2.b Limit <10 gIMi?

All Section 8.0 Acceptance Criteria for this test have been satisfactorily met. Yes_ __ No_

VT-2 Qualified Test Performer * ,,, *$. Date -

Date &-,i-o Date Remarks:-/¢., 1,u,,e' ?2,,Ojao3.Z41/ 4- ..- /oJ- g I _P0,

• • 1 . 0o3% .1to 4 6.o Continuation Sheet Required Yes El No El Page 10 of 13

STP-250.O01A Attachment 11 Page 1 of 2 Revision 2 S #-0TLo k-S _ "

Reactor Coolant System Leakage Test Data Sheet System

Description:

Drawing Number:

Reactor Coolant System See boundary Enclosure(s) listed below. Date: /2 Type of System: Insulated Class: I Type Test System Leakage Test 5.4.1 The Precautions listed Section 2.0 have been discussed with all personnel involved in the performance of the test.

SS_ -,_Signature Date 5.4.2 The required Initial Conditions for this test have been satisfied. 2 ' // IýI/f SS/CRr/US Signature l ate 6.1.2 The piping inside the boundaries listed below is at normal operating pressure for 100% rated reactor power.

6.1.2 oll._ Hold Time Start 6.1.3 OIL)/Hold Time End SS/CRSITUS DATE /

VT-2 Visual.Test Inspected within boundary of enclosure(s). YES NO Remark No.

Boundary Enclosure(s) 3 <C J>f Total Leakage */1,8 (Tech. Spec. 3.4.6.2.lJ(D <1 0 gpm)

All Section 8.0 Acceptance Criteria for this test have been satisfactorily met. Yes_ _ No_

VT-2 Qualified Test Performer _"____---_-- _ Date _2-_-__t_

_ __Date /'-f-'/

v"e- Date J I - -

Remarks:- A //4, ,. - L 'V ,-s Continuation Sheet Required Yes

• No .

Page 11 of 13

I STP-250.001A Page 1 ofAttachment 1I Page 1 of 2 Revision 2 STTS # / 79'/7 Reactor Coolant System Leakage Test Data Sheet System

Description:

Drawing Number:

Reactor Coolant System See boundary Enclosure(s) listed below. Date:,-3, -/1 Type of System: Insulated Class: 1 Type Test: System Leakage Test 5.4.1 The Precautions listed Section 2.0 have been discussed with all personnel involved in the performance of the test.

5.4.2 The required Initial Conditions for this test have been satisfied.

S&/CR*q2ginature Date 6.1.2 The piping inside the bound *es listed bel a ting pressure for 100% rated reactor power.

6.1.2 < Hold Time Start 6.1.3 2-3910 Hold Time End " SVOR ) DATE VT-2 Visual Test Inspected within boundary of enclosure(s). YES NO I Remark No.

Boundary Enclosure(s) _ e _A. _ _ 0' Total Leakage ý0 (Tech. Spec. 3.4.6.2.b Limit <10 gpm)

All Section 8.0 Acceptance Criteria for this test have been satisfactorily met. Yes - No VT-2 Qualified Test Performer -f'ez,. Date jC2M

,Cz,*'*' Date *.~.*

Date Remarks: /A/

&q*" ,/ 3,2ib,57pqb

'. /o3 Continuation Sheet Required Yes E] No 2r Page 12 of 13

STP-250.001A Attachment II RF7V Page i of 2 Revision 2 STTS#//.3/-o*

Reactor Coolant System Leakage Test Data Sheet System

Description:

Drawing Number:

Reactor Coolant System See boundary Enclosure(s) listed below. Data/o- .

Type of System: Insulated Class: I Type Test: System Leakage Test 5.4.1 The Precautions listed Section 2.0 have been discussed with all personnel involved in the performance of the test. II--)z SS/CRS(TUT nature Date 5.4.2 The required Initial Conditions for this test have been satisfied. _,,____-_ / /___-i___

SS/CIRSITUStSiinature Date 6.1.2 The piping inside the boundaries listed below is at normal operating pressure for 100% rated reactor power.

6.1.2 e Hold Time Start 6.1.3 Qfao Hold Time End SS/Ck.f/'YS DATE VT-2 Visual Test Inspected within boundary of enclosure(s). YES [9'*NO I-] Remark No. 1, Boundary Enclosure(s) -3 e - _jc Total Leakage 0 (Tech. Spec. 3.4.6.2.b Limit <10 gpm)

All Section 8.0 Acceptance Criteria for this test have been satisfactorily met. Yes_ _ No VT-2 Qualified Test Performer Date z-*-i,.-

,-/ Date' IA2.-5-/Z.

Date -$'-1.t Remarks: A , t Continuation Sheet Required Yes E No Page 13 of 13