,CATEGORY 3. REGULAT w'i XiVFORMATION DISTRIBUTIO dYSTEM (RIDS) ACCEE/ION 16R: 9907260092 DOC.DATE: 99/07/16 NOTARIZED: NO DOCKET 0 FACIL:STN-'50-528 Palo Verde Nuclear Station, Unit .1, Arizona Publi 05000528 AUTH. NAME AUTHOR AFFILIATION XDE,W.E. Arizona Public Service Co. (formerly Arizona Nuclear Power RECIP.NAME RECIPIENT AFFILIATION Records Management Branch (Document Control Desk) SUBJECT: Forwards rev 2 to "Inservice Insp, Program Summary Manual fo' PVNGS Unit 1," per requirements of 10CFR50.55a(5)(i). C Commitments made by util, listed. A DISTRIBUTION CODE: A047D COPIES RECEIVED:LTR ENCL SIZE:~ Y TITLE: OR Submittal: Inservice/Testing/Relief from ASM Code T NOTES: STANDARDIZED PLANT 05000528 E RECIPIENT XD CODE/NAME LPD4-2 COPIES LTTR ENCL 1 1 RECIPIENT ID CODE/NAME FIELDS,Ã COPIES LTTR ENCL 1 j INTERNAL: ACRS NUDOCS-ABSTRACT RES/DFT/ERAB ;. EXTERNAL: LITCO ANDERSON NRC PDR 1 1 1 1 FILE CENTER RES/DET/MEB NOAC 1 0 1 1 D 'E L„ig L (jQ~) pic-Lu)IiU NOTE TO ALL "RZDS" RECIPIENTS: PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEZVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION '415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 11 ENCL 10
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ACCEE/ION 16R: 9907260092 DOC.DATE: 99/07/16 NOTARIZED: NO DOCKET 0FACIL:STN-'50-528 Palo Verde Nuclear Station, Unit .1, Arizona Publi 05000528
AUTH.NAME AUTHOR AFFILIATIONXDE,W.E. Arizona Public Service Co. (formerly Arizona Nuclear Power
RECIP.NAME RECIPIENT AFFILIATIONRecords Management Branch (Document Control Desk)
SUBJECT: Forwards rev 2 to "Inservice Insp, Program Summary Manual fo'PVNGS Unit 1," per requirements of 10CFR50.55a(5)(i). CCommitments made by util, listed.
ADISTRIBUTION CODE: A047D COPIES RECEIVED:LTR ENCL SIZE:~ YTITLE: OR Submittal: Inservice/Testing/Relief from ASM Code T
NOTES: STANDARDIZED PLANT 05000528 E
RECIPIENTXD CODE/NAME
LPD4-2
COPIESLTTR ENCL
1 1
RECIPIENTID CODE/NAME
FIELDS,Ã
COPIESLTTR ENCL
1 jINTERNAL: ACRS
NUDOCS-ABSTRACTRES/DFT/ERAB
;. EXTERNAL: LITCO ANDERSONNRC PDR
1 11 1
FILE CENTER
RES/DET/MEB
NOAC
1 01 1
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NOTE TO ALL "RZDS" RECIPIENTS:PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTSOR REDUCE THE NUMBER OF COPIES RECEZVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROLDESK (DCD) ON EXTENSION '415-2083
TOTAL NUMBER OF COPIES REQUIRED: LTTR 11 ENCL 10
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CommlfmenL Innmetke. Fnerpy. 10CFR50.55a(g)
Palo Verde NuclearGenerating Station
WilliamE. IdeVice PresidentNuclear Engineering
TEL 602/393-6116FAX 602/393-6077
Mail StatIon 7605P.O. Box 52034Phoenix, AZ 65072-2034
102-04312-WEI/SAB/RKBJuly 16, 1999
U.S. Nuclear Regulatory CommissionATTN: Document Control DeskMail Station P1-37Washington, DC 20555-0001
Dear Sirs:
Subject: Palo Verde Nuclear Generating Station (PVNGS)Unit1Docket No. STN 50-528Unit 1 First 10-Year Interval lnservice Inspection (ISI) Program—Revision 2
Pursuant to the requirements of 10 CFR 50.55a(5)(i), Arizona Public Service Company(APS) hereby submits Revision 2 to the first 10-year interval Inservice Inspection (ISI)Program for PVNGS Unit 1. The enclosed document reflects the final status of ISI
program requirements for the first 10-year interval. Included in this revision is anupdated record of program revisions, revised pages to the program (with change bars),and any new or revised relief requests. Relief Request number 7 was previouslyapproved for limited Class 2 and 3 components, and has been revised to includeadditional components. Relief Requests 9 and 10 were previously approved by theNRC Staff under separate requests subsequent to Revision 1 of the Unit 1 first interval.ISI program. Relief Requests 11 through 16 are new requests being submitted as partof Revision 2 to the ISI Program. Approval of relief request numbers 7, 11, 12, 13, 14,
15 and 16 is requested by July 17, 2000 pursuant to 10 CFR 50.55a(g)(5)(iv).
The following new commitments are being made to the NRC in this letter:
~ APS will perform the required first interval ISI program VT-2 systemleakage test and system pressure test examinations of the ASME Class1 and Class 2 reactor head vent system lines in the first Unit 1 refuelingoutage of the second 10-year ISI interval (U1R8). See Relief Request No.11 ~ p,oQ ~
9907260092 99O716PDR ADQCK 050005288 '' 'DR
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"U.S. Nuclear Regulatory CommissionATTN: Document Control DeskUnit 1 First 10-Year Interval Inservice Inspection (ISI) Program—Revision 2Page 2
Should you have any questions, please contact Scott A. Bauer at (623) 393-5978.
Sincerely,
WEI/SAB/RKB/
Enclosure:
cc: E. W. MerschoffN. KalyanamJ. H. Moorman
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INSERVICE INSPECTIONPROGRAM SUMMARYMAZUAL
PALO VERDENUCLEARGENERATING STATION
UNIT 1
ARIZONAPUBLIC SERVICE COMPANYP.O. Box 52034Phoenix, AZ 85072-2034
3.3.1 The preservice examinations were performed with examination techniques, both automated andmanual, similar to those planned for use during Inservice Inspections. The examinationlimitations noted during the preservice examinations were documented in requests for reliefsubmitted with the preservice examination program. There have been no additional codelimitations noted during the formulation ofthis program other than those contained in theRequest for Relief Section.
3.3.2 All items that are scheduled for examination willbe examined to the extent practical. Inaddition, any code limitations that are noted during the examinations willbe documented in thesummary reports that are prepared aAer each outage.
3.4 EXAMINATIONTECHNIQUES
3.4.1 The three types ofexaminations utilized to perform Inservice Inspections, along with the actualnondestructive examination technique, are identified in the legend below:
PT- Liquid PenetrantMT - Magnetic ParticleET - Eddy Current
VOL- Volumetric
UT - UltrasonicRT - Radiography
3.4.2 Allthe above nondestructive examination techniques willbe performed using specifictechniques and procedures that are identified in ASME Section XI, or alternative examinationsthat are demonstrated to be equivalent or superior to those identified.
3.5 INSPECTION INTERVALS
3.5.1 The Inservice Inspection Program was prepared in accordance with Program B ofASMESection XI. The initial 10-year inspection interval and corresponding inspection periods aredefined below:
01/28/86 to 07/17/98 *01/28/86 to 11/17/9111/18/91 to 03/17/9503/18/95 to 07/17/98
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These dates have been modified to a common interval start date for all three PVNGS units.This is in accordance with NRC letter dated October 21, 1987, from E. A. Licitra, NRC, toE. E. Van Brunt, Jr., ANPP, "Inservice Inspection Programs - Palo Verde, Units 1, 2, and 3"to allow the three units to be under the same ASME Section XIedition and addenda. Itshould be noted that the intervals/periods may change between units to allow for extendedoutage durations per IWA-2400 ofASME Section XI. For Unit 1, a 16-month extensionwas also added to the interval due to the length of the second refueling outage.
* The first inspection interval has been extended 1 year to 07/17/99 as allowed by IWA-2400(c).
3.6 EXAMINATIONCATEGORIES
3.6.1 The examination categories ofASME Section XIwere utilized to develop this program for allsystems, components, and supports. The Program summary tables contained in Sections 4.0and 5.0 are organized by examination category for ASME Class 1 and 2 systems, respectively.For each examination category, these tables identify the system, line number, nondestructiveexamination method, total number of items, required examination amount for each inspectionperiod, and running percentage. For ASME Class 3 systems, the examinations categories areidentified in Section 6.0.
3.7 EVALUATIONAND REPAIR
3.7.1 The evaluation ofall examination results willbe performed in accordance with ASME SectionXIArticles IWAand IWB-3000. In addition, all applicable repairs and replacements willbeperformed in accordance with ASME Section XIArticles IWA,IWB, IWC, IWD, and IWF-4000 and 7000. Pressure tests willbe performed only on welded repairs or replacements, inaccordance with IWA-4000 and 5000. Both the evaluations and repair or replacement willbeperformed in accordance with the 1980 Edition through and including the Winter 1981
Addenda ofASME Section XI, or later editions and addenda ofASME Section XI referencedin 10 CFR 50. Allrepairs and replacements willbe documented in accordance with the WorkControl program, and are maintained at Palo Verde for review.
3.8 SYSTEM PRESSURE TESTS
3.8.1 System pressure tests willbe performed in accordance with ASME Section XIand as identifiedin Sections 4.0, 5.0, and 6.0 for ASME Class 1, 2, and 3, respectively. These tables alsoidentify the type ofpressure test, and test frequency, any applicable requests for relief, andreferences the appropriate ASME Section XIArticle for each ofthe ASME Code Classes.
3.9 AUGMENTEDHIGHENERGY PIPING
3.9.1 Based on the PVNGS UFSAR, an augmented examination is required for protection againstpostulated pipe failures. This augmented examination program includes the followinghighenergy piping systems located between the containment penetration and the main steam
support structure wall:
Main SteamFeedwaterSteam Generator BlowdownDowncomer Feedwater
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3.10 EXEMPTIONS
3.10.1 The exemption criteria identified in the 1980 Edition through and including the Winter 1981Addenda ofASME Section XIwas utilized for all ASME Class 1, 2, and 3 components andsystems. This includes the PVNGS Safety Injection System (RHR, ECCS, and CHR systems)piping and components, even though 10 CFR 50.55a requires the 1974 Edition through andincluding the 1975 Summer Addenda be utilized. Itwas concluded after a detailed review thatthe exemption criteria identified in the Winter 1981 Addenda was more conservative in everycase than those identified in the Summer 1975 Addenda, and more examinations wouldtherefore be performed on safety injection systems piping and components.
3.10.2 A thorough review ofall the systems and components was performed in accordance with theabove exemptions and a complete set ofcolor coded Inservice Inspection Boundary drawingswas prepared. These drawings are maintained at the PVNGS site for review.
3.11 CODE CASES~ e
3.11.1 ASME Section XICode Case acceptability willbe based on Regulatory Guide 1.147.
3.11.1.1 In addition, Code Case N-416-1 was approved for use in an SER attached to NRC letteraddressed to Mr. W. L. Stewart, dated March 16, 1995.
3.11.1.2 Code Case N-498-1 was approved for use in Revision 12 to Regulatory Guide 1.147issued June 10, 1999.
REV. 206-18-99
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0'Q PALOVERDENUCU<ARGE<NERATINGSTATIONM 10YI<~INTERVAL-EXANIINATIONSUMMARYASME CLASS I TABLE 1-6
I- Reactor Vessel Examine the areas aboveand below the reactor corethat arc made accessibleforexamination by removalofcomponents during normalrefueling outagcs.
Requirement The System Pressure Test and System Hydrostatic Test requirementsof the 1980 Edition, Winter 1981 Addenda, IWC-2500, Table IWC-2500-1, Category C-H, All Pressure Retaining Components includeessentially all ASME Class 2 piping.
Alternate Utilize 10CFR50, Appendix J, Leak Rate Testing requirements andresults to satisfy the above ASME Section XI requirements for pipingsystems that pass through the containment wall. This would apply tothe following mechanical/piping penetrations:
The applicable containment piping penetrations are subjected to 10CFR 50, Appendix J, testing. This request for relief (without ASMEClass 2 lines) was originally accepted for the First Ten Year InspectionInterval in USNRC letter dated April 12, 1996, from W. H. Bateman toW. L. Stewart, "Evaluation of the First Ten Year Interval InspectionProgram Plan, Revision 1, and Associated Request for Relief for PaloVerde Nuclear Generating Station, Units 1, 2, and 3". This request forrelief has been revised to include ASME Class 2 piping systems andCode Case N-498-1 for the System Hydrostatic Test requirements. Inaddition, the alternatives of Code Case N-522, Pressure Testing ofContainment Penetration Piping, are being applied for those systemsnot classified ASME outside the containment isolation valves.
For all the listed penetrations, the proposed 10 CFR 50, Appendix J,Leak Rate Testing was conducted at peak calculated containmentpressure, utilizing procedures that willdetect the source and location ofthrough-wall leakage.
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Additional Code Case N-498 is listed as an acceptable case in Regulatory GuideInformation 1.147, Revision 11, and Code Case N-498-1 and Code Case N-522
are listed as acceptable cases in Regulatory Guide 1.147, Revision 12.
On October 28, 1998, Condition Report/Disposition Request (CRDR)number 9-8-1412 was written to address the additional conditionsimposed by NRC Safety Evaluation Report Docket Nos. STN-50-528,STN-50-529 and STN-50-530. It was determined that the condition toperform testing at the peak calculated containment pressure wassatisfied in procedure 73ST-9CL01, Containment Leakage Type "B"
and "C" Testing.
Also it was determined in the evaluation that the additional conditionimposed to perform Appendix J testing using test procedures that willdetect through-wall leakage's in the pipe segments being tested wasalso satisfied. In addition, clarification was added to procedure 73ST-9CL01 for how suspected pressure boundary leakage would beaddressed during performance of Appendix J testing.
Approval In accordance with 10 CFR 50.55a(a)(3)(i), relief is requested from thecode requirements on the basis that the proposed alternative wouldprovide an acceptable level of quality and safety.
1. ASME Section XI, Division 1,1980 Edition, Winter 1981 Addenda.2. ASME Code Case N-498.3. ASME Code Case N-498-1.4. ASME Code Case N-522.
Requirement ASME Section XI, Division 1, 1980 Edition, Winter 1981 Addenda, IWA-5200 and IWC-5220 require the performance of a Hydrostatic PressureTest (1.25 times the system design pressure) after repairs by welding.
AlternateTesting
In lieu of the Hydrostatic Pressure Test required by IWC- 5220, aSystem Inservice Pressure Test (normal operating pressure) per IWC-5221, will be performed.
Basis ForRelief
A detailed basis was provided under separate cover. The following is abrief summary:
~ The hydrostatic pressure test must be performed with the vesseldefueled due to testing prerequisites.
~ This would require an additional reactor reassembly.~ The reactor coolant pumps would be used to raise and maintain
temperatures in a precore type configuration at their upper designlimit, which would put additional stress on the RCPs.
~ All main steam safety valves and atmospheric dump valves wouldrequire removal and replacement with blind flanges.
~ Additional supports would be required due to filing the steam lineswith water.
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Approval Approval granted in NRC letter dated April 8, 1993, from T. R. Quay toW. F. Conway, "Request for Relief from the ASME Code, Section XI,Hydrostatic Pressure Test Requirements".
Examination Category B-A, B-D, B-N-1, B-N-2 and B-N-3
Item Numbers All
Status
Requirement
Approved
The required volumetric and visual examinations of the Reactor Vesselwelds and internals shall be completed during each successiveinspection interval in accordance with ASME Section XI, Division 1,1980 Edition, Winter 1981 Addenda, IWB-2410 and IWB-2500, TableIWB-2412-1. The Inspection Period specified may be decreased orextended by as much as 1 year to enable an inspection to coincide witha plant outage, within the limits of IWA-2400 (c).
AlternateTesting
Reactor Vessel Exams will be performed during the U1R8 Outage.
Basis ForRelief
Delay the First Ten Year Interval Reactor Vessel Exams scheduled forthe 7~ Refueling Outage to the 8~ Refueling Outage in Unit 1.
Reasons for this relief include:
~ The One Year Period Extension for Unit 1 ends 07/17/99 and theU1R8 Outage is scheduled from 10/02/99 to 11/08/99. This willdelaythe exams approximately 114 days at outage completion.
~ This delay would facilitate the Reactor Coolant Pump Shaftreplacements planned for U1R7.
~ The raising of the refuel pool water level for the vessel exams wouldadd approximately 8 days to the 7~ Refueling Outage.
~ There will be fewer critical maintenance activities during U1R8.~ This situation only impacts Unit 1 and does not effect the ISI
Program Plan for Units 2 and 3.
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AdditionalInformation
None
Approval Approval granted in NRC letter dated October 16, 1997 from W. H.Bateman to J. M. Levine, "Evaluation of Request for Deferral of theReactor Vessel Examinations First Ten-Year Inservice InspectionInterval (Relief Request No. 10) for the Palo Verde Nuclear GeneratingStation, Unit 1, TAC No. M99104".
Requirement ASME Section XI, Division 1, 1980 Edition, 1981 Addenda, IWA-5200,IWB-5200 and IWC-5200, Table IWB-2500-1, Table IWC-2500-1 andCode Case N-498, require the performance of a VT-2 Systems LeakageTest (Class 1) and a System Pressure Test (Class 2) at or near the endof the first inspection interval.
AlternateTesting
Defer the performance of the VT-2 System Leakage Test and SystemPressure Test Examinations of the ASME Class 1 and Class 2 ReactorHead Vent System lines to the first refueling outage of the second 10-year interval. In accordance with IWB-2412, the one-year periodextension for Unit 1 ends 07/17/99 and the U1R8 outage is scheduled
.from 10/02/99 to 11/08/99. This will delay the exams approximately 114days beyond the ASME Section XI allowable extension. The deferral ofthese exams will not be credited toward exams for the second interval.The affected portions of the vent lines are as follows:
RC RC-1448
RC-146
ASME Class System Line No.
1 RC RC-179
Line Size P&ID No. Valve No.
RCP-001 HV108to
HV109RCP-001 HV101,
HV102,HV103,HV105HV106
andHV109
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Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested from the coderequirements stated above on the basis that the proposed alternativewould provide an acceptable level of quality and safety.
PVNGS believes that deferring this examination provides an acceptablelevel of quality and safety for the following reasons.
Containment entries are made on no less than a quarterly basis forother plant maintenance. A general inspection for leakage is performedduring those entries per PVNG's procedure 40DP-9ZZ01, ContainmentEntry in Modes 1 Through 4. Furthermore, RCS pressure boundaryleakage is monitored by the control room staff in several additionalways.
1. Containment atmosphere particulate radioactivity monitoring.2. Containment atmosphere gaseous radioactivity monitoring.3. Containment relative humidity monitoring.4. Containment sump level rates of change and discharge monitoring.5. RCS water inventory balance measurements.
Technical Specification 3.4.14, RCS Operation Leakage, allows for only1 gpm unidentified leakage and no pressure boundary leakage. The firstthree methods provide continuous monitoring with alarms. Sump levelsare monitored every hour and the RCS water inventory balance isperformed every three days. Ifgreater than 1 gpm leakage is detected,the leakage must be reduced to within limits within four hours or be inMode 5 within 36 hours.
PVNGS believes that the system pressure monitoring and the severalmethods for detecting RCS leakage provide an adequate level of safetyto justify deferring the pressure test for 114 days. Additionally, theserequired examinations were successfully completed in Units 2 and 3without any abnormal indications noted.
AdditionalInformation
PVNGS is asking for relief to defer the first interval examinations to thefirst refueling outage of the second interval. Due to a long refuelingoutage in Unit 1, Unit 2 has become the lead Unit for implementing theISI Program Plan. Prior to the end of Interval 1 for Unit 2, it wasdetermined that it would require plant evolutions outside of normaloperating practice to perform the required pressure test on the ReactorHead Vent System, so at that time, Relief Request No. 15 was writtenfor Unit 2.
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Subsequent re-evaluation indicated the examination could beaccomplished and the test was performed in Unit 3 during normal plantshutdown. Also, the test has been performed in Unit 2 during the 8~
Refueling and the required test for Unit 1 will be performed during the 8~
Refueling Outage. No remaining scheduled plant shut downs within thefirst inspection interval are scheduled for Unit 1.
Approval In accordance with 10 CFR 50.55a(a)(3)(i), relief is requested from thecode requirements on the basis that the proposed alternative wouldprovide an acceptable level of quality and safety.
Requirement ASME Section XI, Division 1, 1980 Edition, 1981 Addenda, IWC-2500and Table IWC-2500-1 require a volumetric examination of the nozzleinside radius area.
e AlternateTesting
Basis ForRelief
In lieu of the volumetric examination, a surface examination will beperformed on the nozzle inner radius selected for examination.
Due to the design of the PVNGS Steam Generator Main Steam Nozzlesthe volumetric examination is not possible.
AdditionalInformation
The nozzles have a protrusion into the steam generator that makes itimpossible to examine volumetrically. A copy of the nozzle drawing isattached that illustrates these geometric conditions.
Approval In accordance with 10 CFR 50.55a(g)(5)(iv), PVNGS is requesting relieffrom conformance with the above code requirements which have beendetermined to be impractical.
Requirement ASME Section XI, Division 1, 1980 Edition, Winter 1981 Addenda, IWB-2500, Table IWB-2500-1, Category B-A requires the Meridional HeadWelds to be volumetrically examined for the accessible (Note 2) lengthof all welds.Notes: (2) Includes essentially 100% of the weld length.
AlternateTesting
The volumetric examinations of both the Closure Head and BottomHead Meridional welds will be examined to the extent possible. A sketchshowing scan limitations is attached for both areas. The total coverageis estimated to be 31% for the Closure Head welds and 20% for theBottom Head welds based upon past examinations.
Basis ForRelief
These examinations are both limited by physical constraints and currenttechnology. Previous exams performed in Unit 2 and Unit 3 confirm thecoverage listed above and the sketches attached attempt to depict eachlimitation.
AdditionalInformation
The examinations of the Bottom Head Meridional welds are included inthe first ten-year interval Reactor Vessel exams and will be performedduring the 8" Refueling Outage in Unit 1. Deferral of these exams hasbeen approved in relief request number 10.
Approval In accordance with 10 CFR 50.55a(g)(5)(iv), PVNGS is requesting relieffrom conformance with the above code requirements which have beendetermined to be impractical, due to design and geometry limitations.
ASME Section XI, Division 1, 1980 Edition, 1981 Winter Addenda, IWB-2500, IWB-5200, Table IWB-2500-1 and Code Case N-498 require theboundary for the end of interval pressure test be extended to all Class 1
boundaries. This includes the small portion of pipe between two Class 1
isolation valves or between a valve and blind flange.
Visual examinations performed during System Leakage Tests will beextended to include the small portion of pipe and downstream valve orblind flange. The first valve will not be opened. A list of these areas areas follows:
SystemCHCHCHCHCHCHCHRCRCRCRCRCRCRCRC
Line No.CH026CH024CH022CH020CH026CH520CH001RC091RC091RC089RC096RC062RC017RC099RC005
P8 ID No.CHP001CHP001CHP001CHP001CHP001CHP001CHP001CHP001CHP001RCP001RCP001RCP001RCP001RCP001CHP001
The normal reactor pressure boundary is examined during eachrefueling outage and no pressure boundary leakage has been noted.Currently these valves are independently verified closed prior to plantstart-up and are not manipulated during any procedure guided plantevolutions while at power.
AdditionalInformation
PVNGS does not cycle these valves at NOP/NOT because it increasesthe opportunity to experience an incident where a valve will not reseat.This can be due to several mechanisms; foreign material moving intothe seating surface; stem failure while opening or closing; packingshifting; or valve binding. The opportunity for a packing leak could alsopresent itself, with the added challenge of normal RCS pressure behindit. Cycling of these valves and the resulting compensatory actions dueto a leak can easily result in leakage and a forced unit shutdown orcooldown. Current operating procedures require these valves to remainclosed with no exceptions. Valves that need to be operated arespecifically identified to be manipulated only in mode 5 (to prevent RCPseal damage).
The subject piping segments represent vent or drain piping,instrumentation, etc. and are 1" or less in diameter; generally less than12" in length. These piping segments have no active safety function andhave no effect on the operational readiness of the plant.
Approval In accordance with 10 CFR 50.55a(g)(5)(iv), PVNGS is requesting relieffrom conformance with the above code requirements which have beendetermined to be impractical.
Requirement 1. ASME Section XI, Division 1, 1980 Edition, Winter 1981 Addenda,IWB-2500, Table IWB-2500-1 requires 100 percent surfaceexamination of items unless otherwise noted in the code.
AlternateTesting
2. ASME Section XI, Division 1, 1980 Edition, Winter 1981 Addenda,IWF-2500, Table IWF-2500-1 requires 100 percent visualexamination of items unless otherwise noted in the code.
1. A volumetric examination was performed to augment the surfaceexamination, however it was also limited to scans from the skirt sideof the weld only. Both examination techniques were applied to theweld from the outside surface of the pressurizer skirt.
2. The examination of the Spray Pond (Ultimate Heat Sink) pipingsupports was performed underwater, utilizing visually trained andcertified SCUBA divers. This was done in lieu of draining the pondsand cleaning the algae layer off all the welds.
Basis ForRelief
1. Limitations were noted for the Pressurizer Skirt weld due mainly tothe design. The inside surface area of the Pressurizer Skirt isconsidered inaccessible due to the pressurizer heaters,drain/instrumentation lines, and insulation.
2. Due to the environment in the spray ponds, a deposited layer coversa majority of the examination area. A random sampling of weldswere cleaned to reveal any abnormal conditions.
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AdditionalInformation
1. The attached sketch identifies the limitations for both volumetric andsurface exams. It should be noted that the ASME Code requireseither volumetric or surface examinations be performed asapplicable.
2. Cleaning of all the welds was impractical, because this would cloudthe water, making the visual inspection impossible. Ifduring theexaminations, any abnormal indications such as bent, missing orbroken components were discovered, the support would be cleanedto enable a detailed examination of the welds.
Approval In accordance with 10 CFR 50.55a(g)(5)(iv), PVNGS is requesting relieffrom conformance with the above code requirements which have beendetermined to be impractical.
Requirement ASME Section XI, Division 1, 1980 Edition, Winter 1981 Addenda, IWB-5210 and IWB-2500, Table IWB-2500-1, require that Category B-Pcomponents be VT-2 examined at a test pressure not less than thenominal operation pressure associated with 100 percent rated reactorpower.
AlternateTesting
PVNGS will conduct a VT-2 examination on all portions of the reactorvessel, which are accessible during Mode 3, without endangeringpersonnel from undue heat or radiation exposure.
However, in lieu of performing VT-2 exams in areas that are hazardousto personnel, PVNGS will monitor for reactor vessel leakage by the useof leakage detection methods provided in the design and operation ofthe plant.
Basis ForRelief
Pursuant to 10 CFR 50.55a (g)(5)(iv) relief is requested on the basisthat conformance with the code requirement is impractical. Specifically,relief is requested from the requirement to visually inspect the entirereactor vessel while pressurized to the pressure associated with 100percent rated reactor power as required by IWB-5210.
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The requirement to perform a VT-2 examination of the reactor vessel isto detect leakage of the vessel. Because the walls of the ReactorVessel are essentially vertical, the code allows the examination to belimited to the lowest elevation where leakage will accumulate [IWA-5242(a)]. In addition, the code requires that the surrounding areas includingfloor areas be inspected for evidence of leakage [IWA-5242 (b)].
PVNGS cannot comply with this code requirement to perform thisinspection at Mode 3 because of extreme temperature and very highradiation areas.
Temperatures under the vessel during Mode 3 reach approximately 500degrees Fahrenheit. In addition, the exams require personnel to accessareas under the vessel where radiation fields are between 2 to 12 Remper hour.
Accessing the bottom of the reactor vessel while it is depressurized toassess leakage which has accumulated is possible, but the reactorvessel at PVNGS is constructed in such a way that leakage which wouldaccumulate at the bottom of the insulation around the vessel or on thefloor cannot be distinguished from leakage from other sources such asleakage from the pool seals.
While direct visual examination may detect gross leakage, moresensitive methods of detecting leakage from the reactor vessel areavailable, as discussed bellow, which do not endanger plant personnel.
AdditionalInformation
Reactor coolant system (RCS) pressure boundary leakage is monitoredby the control room staff in several different ways:
1. Monitoring of the space between the double 0-ring seal on thereactor vessel closure.
2. Containment atmosphere particulate radioactivity monitoring.3. Containment atmosphere gaseous radioactivity monitoring.4. Containment relative humidity monitoring.5. Containment sump level rates of change and discharge monitoring.6. RCS water inventory balance measurements.
Technical Specification 3.4.14, RCS Operation Leakage, allows for only1 gpm unidentified leakage and no pressure boundary leakage. The firstfour methods provide continuous monitoring with alarms. Sump levelsare monitored every hour and the RCS water inventory balance isPerformed every three days. Ifgreater than 1 gpm leakage is detected,the leakage must be reduced to within limits within four hours or be inMode 5 within 36 hours.
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PVNGS believes that the RCS leakage monitoring performed by thecontrol room staff satisfies the requirement for detection of RCSpressure boundary leakage from the reactor vessel. Performing a VT-2exam on the bottom of the reactor vessel would not provide betterinformation than is possible by other means and does not warrant therisk of injury to plant personnel from extreme heat and very highradiation areas.
Approval Relief is requested in accordance with 10 CFR 50.55a(g)(5)(iv). Thisexamination is impractical due to extreme temperatures and very highradiation in certain areas requiring personnel occupancy in order tocomplete this exam.