ML060190021

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Class 1 Piping Operability Evaluation Submittal Per Code Requirements
ML060190021
Person / Time
Site: Calvert Cliffs Constellation icon.png
Issue date: 01/16/2006
From: Spina J
Constellation Energy Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
05-004R1
Download: ML060190021 (23)
v • d • e


Text

James A. Spina Calvert Cliffs Nuclear Power Plant, Inc.

Vice President 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410.495.4455 410.495.3500 Fax Constellation Energy Generation Group January 16, 2006 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit No. 1; Docket No. 50-317 Class 1 Piping Operability Evaluation Submittal per Code Requirements An anomaly was discovered in an American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Class I Reactor Coolant System line (Field Weld No. I of spool piece l-CC-14) during the review and digitization of original construction weld radiographs by Constellation Energy engineering staff prior to the 2006 Unit I refueling outage at Calvert Cliffs Nuclear Power Plant.

The review of original construction radiographs was performed proactively to provide early review and identification of pre-existing flaws in Class I piping system components.

The initial construction radiograph of the Unit I shutdown cooling outlet nozzle safe-end-to-pipe weld (Field Weld No. 1 of spool piece 1-CC-14) identified a slag inclusion that exceeded the acceptance criteria of American National Standards Institute B31.7 (original construction code). A weld repair was performed during original construction and a follow-up radiographic test indicated that, although reduced in size, the inclusion remained outside of the B31.7 acceptance criteria. Prior to initial plant operations, a pre-service inspection ultrasonic test was performed on Field Weld No. I of spool piece I-CC-14 in accordance with ASME B&PV Code Section XI standards. A second ultrasonic test was performed in 1994, in accordance with ASME B&PV Code Section XI, during an inservice inspection of the same weld. No indications were identified during either Section XI ultrasonic examination.

As a result of the discovery made during review and digitization of the original radiograph, an operability determination was initiated on November 9, 2005, in accordance with site procedures. The initial determination indicated the subject weld was operable but degraded and further evaluation would be required to adequately disposition the indication. A fatigue analysis completed by Structural Integrity Associates, Inc. on November 15, 2005, determined that flaw growth would be small enough to safely allow continued operation for at least two operating cycles.

Calvert Cliffs Nuclear Power Plant intends to perform additional non-destructive examinations of the subject weld during the 2006 Unit I refueling outage, which is planned to start in February 2006. Results of the additional non-destructive examinations are expected to allow more accurate characterization of the radiographic indication. This should allow further refinement of the fatigue analysis by reducing conservatisms included in the initial analysis, thus demonstrating the weld is adequate for the remaining life-of-the-plant. An alternative to leaving the weld in place, as-is, would be to remove and repair the 4°Lf

Document Control Desk January 16, 2006 Page 2 weld during the 2006 or 2008 Unit I refueling outage. This option would be used only if additional characterization determines that the flaw growth is unacceptable for the remaining life-of-the-plant.

The attached operability determination is provided for Nuclear Regulatory Commission review and approval in accordance with ASME B&PV Code Section XI requirements contained in IWB-3640, "Evaluation Procedures and Acceptance Criteria for Austenitic Piping." Calvert Cliffs Nuclear Power Plant intends to provide Nuclear Regulatory Commission with additional evaluation results after completion of the Unit 1 2006 refueling outage, but in no case later than the start of the Unit 1 2008 refueling outage.

Should you have questions regarding this matter, please contact Mr. L. S. Larragoite at (410) 495-4922.

Very truly yours, for James A. Spina Vice President - Calvert Cliffs Nuclear Power Plant GV/MJY/bjd

Attachment:

(1) Calvert Cliffs Nuclear Power Plant Operability Determination No. 05-004RI cc: P. D. Milano, NRC Resident Inspector, NRC S. J. Collins, NRC R. I. McLean, DNR

ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT OPERABILITY DETERMINATION NO. 05-004R1 Calvert Cliffs Nuclear Power Plant, Inc.

January 16, 2006

ATTACiIIENTI' 2, DETERNIINAiiON FOR FORTEACIIHI'V ThCII SPEC SS(CS (PAGh I OI 3)

01) NO.:05-004°Z I l)A'I'Iifl'IM' IN I'I'IA\'l'IiI): I.-/15/05/12:21 (Saiule 01) DImil,,e, lsed oil Attachient 7)

UNIT: I ISSJ1 'UEIORTII#: IRE-009-389 I.QU 1PMlN'l/C()OI\'l( )N.N'i' 1)l'SCRI IIOI(N: (S; I;'1'l lti///( '0vl~l///UlJI#//E'1('. )052

( PER~lA I§II .ITI' IRlEC'(NINI\IINI)ATI'ION ('IEI5('I.ISTI C(IIhECK ONE O 't'i1I'OI.IMW1NC:

I . 3 The affected stIrIUCtuIr-e/systeCnIicoim)oinent (SSC) should be deClareCd OlERIAlLEI s reasonablC assurance exists which indicates that tIhe dlegriadICdinoii-conIformiing SSC WILL P1'R FOR lvi its inten(ded salfly Iunctlon(s) as required(.

2. O The aftctCed stIrictiie/Systeiii/coi1lpOi1nnt (SSC) should be declared INOPERABLE as reasonable aSSUranice of the S(' Functionality D0lES NO(-' exist and(l Ie (leSgrade(l/n1OIl-coiifoi-IniLig SSC' WILl. N' TPERFORM its intended safety Iunction(s) wvIieII required. 'ITerniinat( the Ise ofI his attachment and lilimcdiately iill;rni n1le (GS - NPO or Shift Manager of the inoI)erability.

I)OC('lUENTA'I'ION O OPE'RABILITY RECOMNMIENI)ATION

)escription of the issuIC/situatlioll (tllat resulted ill the nlecd for til lunctional IEvaluation):

While pertormiing 2006 reftileing oulage preparations, original construction weld radiograpl information was rev'iew'ed. D)uring thle review, it was determined that Field WeldI # I in thle class I Shlit l)oWn Cooling (Sl)C) line has a non metallic inclusion that should not have been accepted dUIring COInStruICtlOI. (Construction co(e 133 1.7 Appendix 13-1-140, 1969) dition allows the acceptance of inclusions vithl anl aggregate length no greater than thle thickness of the pipe whlcin there is a line of inclusions. In this case, there are a group of tiree small inclusions with a combined lengti of 1.5". BecaLIsc thle thickness otfthe pipe is 1.125"n the inclusions should nlot ha1ve beein accepted during thle origillal analysis of thle radiograph.

At(er acceptance of thme construction weld, aI Pe-Service Inspection (I .l) was completed on thie weld. T'his was comiducted in accordance w;'ith (lie requirements of ASM Ii Section XI, 1970 Ed, Summer 1970 Addenda. TIhis section directed tihe use ofl ASM E.Section Ill, 1970 Editionl Summer 1970 Addenda. The examination examined essentially tile full Voluime of the weld, using calit-ration blocks with 1/4t, I/2t, and 3/4t side drilled holes. 'I'llis exam (lid not detect the presence oh 'ally recordable md ications.

An ASME Section Xl exam wvas satisfactorily completed in 1994. This exam was completed in accordance with ASMI Section Xi, 1983 lid, Summer 1983 Addenda. IThe exam did not detect the presence of any weld inclusions. Ihe exam volume of thle ASMEl Section Xi weld exam is tile inner 1/3 thickness ofthe weld. TIhis examll Was conducted fIrom rboth sides of tle weld. While tihe Section Xl exam concentrated oIn thIe inner 1/3 of the 'eld,time exam technique used an extended beam path that examined, essentially, the Full v'oluime ofthe weld.

'Ihe inclusion could be OLItSide of'tlis VOIlnIIne Or it COuld lbe aligned sucII that tile angle beaImls used in the ASMIE Section Xl llIvwere not abile lo detect tile inclusion.

NO- I -1(06, Revision 10 I nollorms/ I- I06-02.dot

2. Impact onl Nuclear satiety and operation (ID)escribe the potential or actual impact of the issuie/SituationI onl nu.lclear salfety alnd olpCerations):

I)ue to the characteristics of tlic indication, there is no immeinilate impact onl Nuclear Safllty.

B3ecause the inclusion was bound (lurIinlg thle colstrIuction weld RT, it is not a service induced flaw. Tlhere are three potential causes which could influence growth:

1. Pressurized Water Stress Corrosion Cracking (I'WSCC)
2. Inter Granular Stress Corrosion Cracking (IG(SCC')
3. Fatigue Because [l ie PSI undtlhe 1994 ASMFI exams did not identify ally indications at tlie 11) of lthe pipe, l'WS(!C and IGSCC call he eliminated as potential phienomenon which would cause the flaw to grow. Without contact with tile pumnped fluid, conditions are not available to induce this type of flaw growth.

Structural Integrity Associates, Inc. (SI), perlformed Fatigue analysis using conservative assumptions and determined that the flaw growth wotuld he slow enough to allow COlitintle(

operation for at least two operating cycles (until 2()0 (1 0 FO) wvith margin. 'The analysis is based o0n tlhe assumption that:

I The

[. inIclusionI is a 1.5" circuniiCereintial lla\\

2. T'he flaw initiates at the pipe Of) (a satisfictory 1I' xamination was perltorined on the outside ol' the pipe during tIle 1994 VOlUmrretric examination)
3. TIhe flaw-extends 66',', through tile vall of the pipe (based on tile satisfactory doCumented results of0Ithe inner 1/3 during tlhe 1994 v'olumietric examination)
3. Regulatory relutiremeilniiits/collllitlllelnts()escribe thle potelitial or actual impact of tile issic/

siluation ol e ('tleirrent license Basis):

Acceptance standfard(s for welds are identi lied in l\VB-3 13 1, which references table IW\I-34 10-1, which states the acceptance standards o01 IWB 3514 arc to the u.ppfiec0.

or the assumptions placed onl this f1aw, thle acceptance criteria of table 1\13-34 I-1-I are not met. IWI3-3131 goes on1 to require tile condition to be corrected under IW3B-3132.2 or IVB-3132.3 IWI3-3132.3 permits aln analytical evaluation of thle flaw as ijn IW13-360t0.

IWB-364() contains the evluiationl proced urcs and acceptanICce criteria for aItSnCitic I)ililng. SI perflormed tihe analysis to these procedtires and criteria and found the flaw to lc acceptable F'or at least 4 more years of service, Or 2 more operating cycles, b)eyond the current operating cycle. In accordance with IWI3-3640, tile evaluation procedures and acceptance criteria shall be the responsibility of0the owner and slall be subject to approval by thle regulator)y authority having jurisdiction at the plant site. IB;ased on this requirement, CCNPP must obtain NlR(

apl)roval of'the Si evaluation beflore restoring tlhe system to flull qualification and c losing this NO-l -106.

NO-1-106, Revision 10 nol'orins/ I- IO6-02.dot

Given the need for NRC approval of the SI evaluation, CCNIIP shall remain in, 'J'RM 15.4.3, Structural integrity ol ASMLE C(o'de class 1, 2 iind 3 components shall be within the limits of the In-service Inspection Program, until the NRC approves the SI ev'aluation.

Tech Spec 3.4.13, RCS operational leakage shall be limited to no pressure boundary leakage.

The system is in compliance with this 'Iech Spec. ']here has been no increasc in ICS leakage and the bare metal visual inspection of a weld in close proximity during the 200-I4 RO revealed no indication of leakage.

4. Stiructtiue/Systemii/Componenit (SSC) saflety function(s) (Fully describe the SSC sallty functions, particularly those that are potentially impacted due to the issue/sitluation):

The safety Function of the affected pipe is to provide a qualified pressure boundary connection between the Reactor Coolant System (IZCS) and the Sl)C system and an RCS pressure boundary to the containment atmosphere while at power.

Revision Number:_

Basis for the Revision:

NO-I-106, Revision 10 notorn-s/l - I 06-02.dot

AT'T"I'A(.'l INIE:NT 2, ()FORI I'J ' I)l:'l INA'IiON I'()It 'IlC'i(ll iSPEC ( 's (Page 2 olt3)

5. Evaluation:

A. Scope of'eval natioI:

Ihlis evaluation will restore the SySlem lo hill[ qualification through 29()10.

B. Applicable specific eCvents and scenarlios (associated with the issue/situa.ltioll):

Tlis calLiation wvill apply to all mo(les of operation when the aflveted SIl) line is required to be operable.

C'. Givens/assuimptions (information that supports tIle specific evaluation, inclllding adverse impact):

For thle pLurpose o tllis evaluLation, tlle 11olowing assutmptions were conservatively illade.

1. I[he l.5" long circumillerential weld 1lamv is 66%' through the wall thickness fi-on tle 01) of the pipe.

IPast inspections of this weld include a post construciioll PSI. This WaS essentially a full volume inspection of tlie weld that did not (delect thle preseicc o ally recordable indications. More recently, tile inner 1/3 thickness of the weld was cxamined in 1994. Again, tile inspection (lid not detect the presence of any weld inclusions. B3ased oin these twvo examinations, it was conservatively assumed that tlhe weld flaw is 66%',`

(0.7425") through the wvall thiekness firom tile 01) of tile pipe.

2. The flaw is connecte(l to tlhe 01) of the pipe lI)u-inlg tlhe 1994 inspection,.a PT1 inspection of the \clld was perlohrned with no indications identi fied. Conservatively, this evaluation wVill assume that thelie 1l\ is connected to tie 01) of Iile pipe.

Until tlhe issue with this weld is resolved, Operations will continue to monitor R(C'S leak late fIr increasing trends and ideiitily tile SOUfICC IAW the guidance in Operation's Standing Order 03-03 IW('S Leakage".

1). Speciflic evaluations (D)ocument tIle results and long-termn capabilities of'lthc SSC):

Si has performed an analysis of'lic tpropagation of' a fatigue Ilaw under thle conservative asstimIf)tlions outline(l above. 'heir analysis (see attached) indicates that flaw growvth would he relatively modest such thalt the flaw\\, can be shown to meet ASFIF Section XIl allowable flav Criteria folr at least two more operating cycles (2010(ZFO) with margill.

F. Method lo restore SSC (e.g.: repair, Mod) (I )ocument only the intended actions to restore tlie SSC to flull (ualification)

IAW RVlI-3640, NRC approval of' tIhe Si evaluation %villhe re(qllire(l to restore tile system to 1f1ll (ILfalification thlroLIgh 2010.

NO)- I - 106, Revision I () nol'orms/1 -I 06-02.dot

F. Estimated Completion D)ate (1[C'I)) (For each action):

  • (O)htain NRC approval of the Si evaluation - IR200500308 ins 003 - TIBD)
  • ID)evelop the inspection p)lan to 1be perl'ored during the 2006 RC) 1R200500308 ins 004 - 12/16/05
  • Identify additional corrective actions flollowing the 2006 RE() inspection I R200500308 Ills 005 - 5/26/06
  • Implement correctivwe actions flowloing the insl)ection evalIuation I- 2005(J0308 ins 006 - TB131)

(. Exp)cted Iplllt Conf'igulration including the ellect ol('onipensatory Acltiols (I)ocument the safest plant configuration incLtiding the lcuedt ol'any transitit la; actoills).'

Based on the stated operaliility ol'lthe veld, it is accetltable to maintain normal plant operations. Operations wvill continue to monitor RC('S leak rate Imr increasing trends and identify the source lAW the guidance in Operation's Standing Order 03-03 "RICS I eakage".

Because the uinit is required to he shut down to p)erform inspections onl lihc weld, the 2006 IW0 will he uiizi'Led to perl brn1 il inspections outlined iti section 6.

Include aniiy spec iaI iiIctihods or philiti conitlioitlis needed to pe per It iiSn lyveit hiicc test iig to Ia idih2ain opcrab)ility. 1B04961 No-I - I 06, Revision I() nlot;brils/ I - I 00(-029.dot

AT'llACllHMENTI' 2, ()FlsRTlVIIT I)IC'IERNIINA'I'IIN FOR 'lL'(I SPCS(' SSCs (Page 3 ol 3)

6. Recommeiidations 1or 1utill her evaluation (\liy sIhloLild it he considered):

l)etermine an evaluation plan to inspect tlie weld during the 2000 RFO. BaLse(d oil the ASME Section Xl inspections that have bcein perloried to date and tile SI anllysis that colnservativCly confirmis continued operability uintil the 20 10 R1l(),

tile Weld is considlered operal)le at this lime. Io substantiate tins position, th1e wel(l in question wvill he interrogated closely during tile 2006 RFO.

Ihis inspection will:

1. Verify/locate the ilcluLsion1 docUllmcntcd onl t0le original collnstrctio radiograph inspection
2. Validate the assumIpt)Iionls useed in the SI analysis
3. I)eterrmine if repairs are necessary
  • Identity additional corrective actions required following the 2006 IZl'O interrogation of the subject weld. This wvill determine il' the Weld muLsIt he repailCd or can be accepted as is.
  • Implement corrective actions heVelop)eId (ILIh illg the evaluationilo the 2006 RFO illspection results
7. References (Supports the specific evaluation):

1 SI analysis CA06657

2. ASMEI Section XI, 1998 Edit ion, Section IVII-3 112(a)
8. Attachments (Applicable items in Step 7):
1. SI analysis ('A06657
2. 1Rl1-009-389
9. EiqUipment is (Check One):

O1TRA1BLE 2 / INOPEhkAlHII I.

Prepare / _~ .i'

~/ivilt~l'eDate

~SI 1, ime Rexviexv ed by: 1? ,"- d/ ///

Signa~tuie D~ate / .

Approved by GS-PIES:. . __ _

(or designee) Signatlul DI)ae lime Rccommendation is (C'ieck One):

ACC('II'IT) II' REIJE (CI'IED) ° (iS-N PO: c _ _ __

/ !ll  :'

(or designee) Signatulre 1)Date If Recommendation is IUJEC'T'1D, provide reasons below:

A'I' No.: IR No.:

NO- I - 106, Revision (0 I nol'ornis/ I - 100-02.dot

1(. Inactive Operability D)etcrmination (iGs-PES (or1 desiglnce):_ _ _,,_ _____

(;S-NP() (or (lesignee): -- -I_ -__-

Original l o: ( ontrol toomi's Active ltIlnctionIl 1hvIltIatIioII/( ))Cl-rIi Iity I )eter li 1 1 1ion IIo0ok.

poll Compll)letion, i 1no aclion is to he Iakeni, then process this Attachmicnt per Sectioni 7.0.

NO- I - I 06, Revision I n) 110 f~lorls! I - I 06(-029.dott

1110120no5 Page 1 of 2

=

I CR # IRE-009-389 1 CONDITION REPORT MOU:

I.

Do you have PersonnelEquipmrenl Safety Concern? [ N] 2. Do you have an Operability Concern? 3. Do you have a Reportability Concern?

I

,4.Do you have a iPotenbal Trip or Reactvity Concern? [ 3 5. Should the area/equipment be Ouarantined? [N] Additional Information Attached? 1L ReIaed?

elate, q

[RO

6. Condition Descr. IHILE REVIEWING AND DIGITIZING TNE ORIGINAL CONSTRUCTION RADIOGRAPHIC FILM TO SUPPORT ISI. AND DISSIMILAR METAL WELD XAMINATIONS DURING THE 2006 OUTAGE REVEALED AN INDICATION THAT DID NOT MEET CODE. BECHTEL WELD NUMBER 1 ON DRAWING 1-23-10, LINE CC-14 HAS A SLAG INCLUSION THAT EXCEEDS THE ACCEPTANCE CRITERIA INANSI B31.7 (ORIGINAL CONSTRUCTION CODE). THE INITIAL NDICATION WAS IDENTIFIED AS SLAG. AND THE WELD WAS REPAIRED. THE RE-SHOr OF THE REPAIR (RI) SHOWS A REDUCTION INTHE LEtJGTH OF

'HE INITIAL REJECTED CONDITION. BUT DID NOT REDUCE THE SIZE TO AN ACCEPTABLE LENGTH. THE RADIOGRAPHIC VIEW ISIDENTIFIED AS 9-16.

JITH A LENGTH OF 1.78 INChES.

7. Date!Time Discovered: lN10312005 6 8. Activity In Progress when discovered: JRADIOGRAPI fiC FILM REVIEW AND DIGITIZATION
9. Immediate Action Taken: INFORMED SUPERVISION.GENERATED CR. DISCUSSED WfIH CODE KNOWLEDGEABLE PERSONNEL J 10 Is this a Recurrirg Condition? L!
11. Apparent Cause: NTERPRETATION ERROR DURING REVIEW 12. Extent of Condition: pggNWN
13. Recornmended Actions: pETERINE FCODE COMPLLANCE ISSUE EXIST. SEVERAL ISI EXAMINATIOtNS HAVE BEEN PERFORMED ON TNIS WELD AtD ACCEPTED.

HARDWAREINFORMATION 14.Unit#: [T l5.EqpLc. [l 16.VendotlMlg l

17. UlEls: ________EpL__16__Veno___M__

lSYS052 - SAFETY INJECTION 052 SYSTEM 19.Equipstatus: N ItEquipmentisstdli [-l Placed? lZJITiiieiiiriZ

. j F Location: [ J OOS? servIce.isit Degraded? IJ NOH-lARDWARE INFORMATION 20. Related Documents: [

21. Initiatof's Name: pEED. ALVIN S I Ext 95-2089 ] Group: li 03 I

l Date: 105 l Time:

(wil

-9 AT lte 14 * .241 0*oe 1.Isthis an Immediate PersonneU/Equipment Concern?

Li [ RECO attached [u Condition could. but does not affect operability of SSC

2. Do you have an Operability concern inany Mode?

m- [q Condition made SSC Inoperable. but operability restored Do you have a Reporlability Concern?

E1 [I Condition COULD NOT affect operability of an SSC L I

4. Do you have a Potential Tnip or Reactivity Concern? Operability Explanation: FIPE INQUESTION (SDC) IS DESIGNED TO 94.5% OF THE ALLOWABLE VALUE Ml tOR DBE LOADING. THE CODE PROVIDES A SF OF 4:1. iSt EXAMINED IN1994
5. Do you have a Plant Tampering Concern? hND ACCEPTABLE FOR ASME SECTION Xi. WILL NEED TO ACCEPT AS IS OR

)EMOVE THE FLAW PRIOR TO RESTART FROM 2006 RFO.

6. Recommended Categoy [l 7. MO Reromtended? [ Should an outgoing OE be issued?
9. Fitness for Duty Evaluation Considered? 1+/-1 10. Compensatory Actions Taken: l
11. Was Condition Corected on tie spot? LI] 12. Recommended Group to Resolve CR: l 13. Discussed withl:

L S Resonrmnended Group to A1003 14.Arefurtheracdonsrequired?

LJ 15.

Renbotve Programmatic CR: 1 1 riNGINEERING PROGRAMS

16. Special Indicalors Assigned:
17. Recommended Actions to Resolve Cond: CONDUCT ADDITIONAL NOE DURING 2006 RFO, EVALUATE FLAW. ACCEPT AS IS OR REPAIR TO MEET ACCEPTANCE CRITERIA.

1B. CR Approved? [ l Name: lIMOTHY LUPOLD I Phone: l283 - Approved Date:

1:7.1'j EW161 s II*five Q itFAC14' .4 NATJ 11AY

1. Isthis an Immediate Personnet/Equipmenl Safety Concern? [Ii 2. Is this an Operability Concern in the current Mode? [W]
3. Isthis a Reportability Concern? RM 1-101 report # [I [i ] 2a. T.S. # 1.43 -
4. Isthis a Trip or Reactivity Concern? [] 2b. This would be an Operability Concern in Mode: O 5.Operabilty Determination Implemented per NO-1-106? [ 6. Compensatory Actions Taken:
7. Comments: rESTORE PRIOR TO EXITING 2006 RFO.
8. Name: AYGAIES DatelTime: l 1700 Phone:

me31I W15 = -

10 Required? L 2. Priorit' [y 2 ] 3. Work Type: [ 3 4. Mode to Work [ii 5. RMG: E

6. IsCR Programmatic? j 7. Mode Restraint [I] Mode Code: I 8. Shift Manager; r appgi Requijed print k)stbrtirg wok? ]

I I I I .

EN-1-100 Fom-s Appendix RevisionI 4 ESP No.: ES200500643 Supp No. 000 Rev. No. 0000 Page 1 of 12 FORMI 19, CALCULATION COVER SHIEET A. INITIATION (Control Doc Type - DCALC) Page I of 12 DCALC No.: CA066S7 Revision No.: 0000 Vendor Calculation (Check one):

  • Yes D1 No Responsible Group: Mechanical & Civil Engineering Unit Responsible Engineer: Andre S. Drake B. CALCULATION ENGINEERING [ civil [ Instr& Controls ] Nuc Engrg DISCIPLINE: Electrical
  • Mechanical E Nuc F'uel Mugnit O Other: El Reliability Engrg

Title:

Prediction of potential crack growth of weld indication.

Unit 1 2O COMMION Proprietary or Safeguards Calculation 2 YES

  • NO Comments: This calculation is for resolution of IRE-009-389 which identified a weld indication on the Shutdown Cooling Outlet Nozzle Safe-End-to-Pipe weld.

Vendor Calc No.: CCNP-06Q-301 REVISION No.: 0 Vendor Name: Structural Integrity Associa ites Safety Class (Check one): a SR 0 AQ El NSR There are assumptions that requnie VerificatiOll durifig walkdown: Al l fl:

This calculation SUPERSED)ES: N/A C. REVIEW AND APPROVAL:

Responsible Engineer: Structural Integrity Associates 11/15/05 Printed Name and Signature Date Owner Acceptance Andre S. Drake n / 19 a4J 11/15/05 Printed Name and Signature Date Approval: Jack J. McHale ItII 5a it S" PrintV Name and Signature / / Date IF the results or conclusions of this calculation or revision mioht affect a procedure or the basis of a procedure, a Change Notification Form (Form 14) shall be forwarded to the Procedure Development Unit with a suimmnary of the calculation's purpose and results.

13GE Calculation Number CA06657 Revision )0()()0 P~age 2 last of [Flflective PaLes Page No. Revision 1I 2 0 Appendix 1: StLutIC.ual Integrity Calculation No. CC(NP'-06Q-30 1, "Prediction ol 'otential Crack Growth Rate of Weld Indication ounIld in tlhe Unit I ShIltdown1 Cooling Outlet Nozzle Safe-End-to-Pilpe Weld," Rev. 0.

'Table of Contents Page No.

Calculatioll Coverslheet ......................................... l Effective Pagesrfable of Contents Reviewer Comments ....... 2 Appendix I - Structural Integrity Calculakltion No. CCNI'-06Q-30(1, Rev. 0. (10 pages).

Reviewer Co('0uments.

1. In Section 2.3 it is mentioned that radiograplis were taken in 1994. It is clarified that thle NDF method employed in 1994 was an ultrasonic examination. This does not impact the computations, nmetlhodology, or concl usiolIs of this analysis.

APre t&Il I 4v CAo 665-7 (ID PC . t5 )

VAssociates, Structural Integrity Inc.

CALCULATION PACKAGE File No.: CCNI-06Q-301 Project No.: CCNP'-06Q PROJECT NAME: Shutdown Cooling Outlet Nozzle Safe End-to-Pipe Weld Indication Evaluation Contract No.: 416596 CLIENT: Constellation Energy PLANT: Calvert Cliffs Unit I CALCULATION TITLE: Evaluation of the Shutdown Cooling Outlet Nozzle Safe End-to-Pipe Weld Indication at Calvert Cliffs Unit 1 I

Project Mgr. Preparer(s) &

Document Affected Reso ecito prvlChxecker(s)

Revision Pages Signature & Signatures &

Date Date 0 1-10 Original Issue < (P)W Computer I 1/( o)

Files A ¢jt CJ°)

.'57C PgCf l E'agelI of 10 SI Form F2001R2a

.I Table of Contents I INTRODUCTION ............................................. 3 2 TECHNICAL APPROACH OR METHODOLOGY .......... .................................. 3 2.1 Indication Size ............................................. 3 2.2 Allowable Flaw Size ........................... I3........... 3 2.3 Crack Growth ............................................. 4 3 ASSUMPTIONS / DESIGN INPUTS ............................... 4 3.1 Allowable Flaw Size Design Inputs/Assumptions .4 3.2 Crack Growth Analysis Design Inputs/Assumptions .4 4 CALCULATIONS .. 4 4.1 Allowable Flaw Size .4 4.2 Fatigue Crack Growth .5 5 RESULTS OF ANALYSIS .. 6 6 CONCLUSIONS AND DISCUSSIONS .. 6 7 REFERENCES .. 10 List of Tables Table 1: Plant Transient Condition [7] .7 Table 2: Bending Stresses due to Pressure and Gravity .7 List of Figures Figure 1: Fatigue Crack Growth .9 Structulral Integrity File No.: : CCNP-06Q-301 I Revision: 0 V Associates, Inc, Page 2 of 10

11I INTRODUCTION Based on information provided by Calvert Cliffs in References 1 and 2, during a recent review of the construction radiographs of the Unit 1 Shutdown Cooling System outlet nozzle safe end-to-pipe weld, a non-metallic inclusion was discovered. The inclusion consists of three closely spaced circumferential inclusions with a total length of 1.5 inches. Based on a 1994 ultrasonic examination of the weld, the inside third of the wall thickness was indication free. The end of the indication (nearer to the outside surface) could not be confirmed based on the available information. The insulation was removed from this weld during the 2004 refueling outage in support of a bare metal visual inspection of the dissimilar metal weld located in close proximity. This provided the opportunity to identify if the weld was leaking. There was no leakage identified when the dissimilar metal weld bare metal visual inspection occurred. Since a surface examination of the weld was not performed, there is no inforuation available to detenninie if the inclusion observed on the radiograph is connected to the outside surface.

The indication was evaluated using the acceptance standards of the ASME Code [3]. It was concluded that the indication dimensions did not meet the acceptance standard in IWB-3500. The indication evaluation was therefore performed to the requirements of IWB-3600 of the ASME Code. Specifically, since the safe end, connected elbow and weld materials are stainless steels, the provisions of IWB-3640 of the ASME Code were used to perform the evaluation. The details of the evaluation and its conclusions are provided below. It should be noted that the indication is in stainless steel material and it is not connected to the inside surface of the pipe. Therefore, the only mechanism that requires consideration is fatigue.

2 TECHNICAL APPROACH OR METHODOLOGY The acceptance of the indication in the as-is condition requires consideration of potential crack growth, applied stresses, and allowable flaw size. The allowable flaw sizes incorporate the required safety factors per ASME Code,Section XI, IWB-3640.

2.1 Indication Size The depth of the indication was found to be 0.7425 inches, or 66% of the wall thickiess from the outside surface, as discussed in Section 1. The length of the indication was determined to be 1.5 inches. The pipe thickness at the indication location is 1.125 inches and the inside diameter is 10.5 inches (12" Schedule 140) [4].

2.2 Allowable Flaw Size ASME Code,Section XI provides acceptance criteria for flaws in austenitic piping. Tables IWB-3641 -1 and IWB-3641-2 provide allowable end-of-evaluation period flaw depths for normal and emergency/faulted conditions, respectively. Table IWB-3641-1 of the ASME Code,Section XI gives the allowable depths as a function of stress ratios and the ratio of the flaw length to the pipe circumference.

Section 4.1 provides the allowable flaw size calculations.

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2.3 Crack Growth Based on the information provided in References 1 and 2, it is believed that the indication is an original fabrication-related subsurface defect (non-service induced) that could potentially have broken the surface during operation. Potential crack growth mechanisms include stress corrosion cracking (SCC) and fatigue. SCC can be attributed to primary water stress corrosion cracking (PWSCC) or intergranular stress corrosion cracking (IGSCC). PWSCC is not a concern here because stainless steels have been shown to be resistant to PWSCC and the indication is not exposed to the coolant. IGSCC has typically been a problem for the boiling water reactors (BWRs) and has not been a concern for the PWRS due to reduced levels of oxygen in the primary loop. Since this location is subject to thermal cycling, however, crack growth from the time when the radiographs were taken (1994) must be considered.

The fatigue crack growth calculations are presented in Section 4.2.

3 ASSUMPTIONS / DESIGN INPUTS 3.1 Allowable Flaw Size Design Inputs/Assumptions

  • Normal Operating Pressure: 2.235 ksi [4]
  • Design Temperature: 650 'F [4J Moments at weld are provided in Reference 5
  • Indication Size per Section 2.1
  • Design Stress Intensity: 16.7 ksi (A376, Type 316) [6]

l SMAW or SAW field weld

  • Indication is connected to outside surface 3.2 Crack Growth Analysis Design Inputs/Assumptionus
  • Indication is connected to the outside surface o* Fatigue crack growth is due to systein thermal and pressure cycling 4 CALCULATIONS 4.1 Allowable Flaw Size The applicable stress range formula for SMALW or SAW field welds to input to Table IWB-3641-1 and IWB-3641-2 for allowable flaw size for circumferential flaws is:

Stress Ratio = ZP+- 2.77]

Where:

Z = 1.15 [I + 0.013 (D-4)] forSMAW

= 1.30 [1 + 0.010 (D-4)] for SAW SStructuralIntegrity File No.: : CCNP-06Q-301 Revision: 0 V- Associates, Inc. Page 4of 10

Pm = primary longitudinal membrane stress (P*IJ(2t)), ksi Pb = primary bending stress (D/(21)*Resultant moment at weld), ksi Sm= Allowable design stress intensity Pe = expansion stress resulting from restraint of free end displacement, ksi D nominal outside diameter of the pipe, in.

d nominal inside diameter of the pipe, in.

P operating pressure, ksi R - nominal outside radius of the pipe, in.

I = moment of inertia (rt/64*(D 4 -d4 )), in.4 t = nominal thickness, in.

Pb includes bending stresses due to dead weight plus operating basis earthquake (OBE) loads for normal/upset conditions and dead weight plus design basis earthquake (DBE) loads for emergency and faulted condition.

Pe includes bending stresses due to plant heat-up.

Substituting in the above equations yields a stress ratio of 0.89 for normal/upset conditions and 1.14 for emergency/faulted conditions for an SAW weld (SAW results in worst case Z). Note that primary bending stress for the emergency/faulted condition is conservatively assumed to be twice the stress for the normal/upset condition. For the observed indication, the length to pipe circumference ratio is less than 0.1. For these parameters, the allowable depth is 75% of the pipe wall. The actual depth from the pipe outside surface is 0.7425/1.125 = 66%.

The calculation details are provided in the project files.

4.2 Fatigue Crack Growth Fatigue crack growth for two additional cycles was done on indication using pc-CRLACK%software [8],

and TS-2 software [9]. TS-2 calculates the thermal stress at the local section due to thermal transients.

Table 1 shows the transients considered for the fatigue crack growth analysis. Bending moments for pressure and dead weight (DW) during heat-up were extracted at the location of the indication per Reference 5. Using the bending moments for this location, bending stresses due to internal pressure and DW were calculated as follows:

Bending Stress due to pressure + DW = My *.zl = 6.596 ksi Since Pressure = 2250 psi at end of the heat-up transient Hoop stress = Prmcan/2t = 5.812 ksi Bending stress due to DW = 6.596 - 5.812 = 0.784 ksi Hoop stresses during all other transients listed in Table I will be factored based oln the operating pressure at each transient [7] and the above hoop stress calculation. Bending stress due to DW for all other transients listed in Table I will remain the same.

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The axial stress distribution from OD to ID of tlhe nozzle safe end at the indication location for several transient conditions was calculated using the TS-2 software [9]. The transients discussed below are the only significant transients with respect to this analysis and were used as the basis for calculating the stress response of all transients.

1. Heat-up: This transient was used to establish the baseline steady state case.
2. Loss of Secondary Pressure: Additional load case due to rapid temperature change. This load case was included in addition to the factored steady state load case during Loss of Secondary Pressure Transient stated above. The corresponding pressure stress variation wvas also included.
3. Reactor Trip/Loss of Reactor Coolant/Loss of Turbine Generator Load: Additional load cases due to rapid temperature change. These load cases were included in addition to the factored steady state load case during each transient stated above.
4. The steady states for all other thermal transient stresses were determined by applying a factor to the steady state of the heat-up transient.

All transients with the appropriate scaling factor at the beginning and the end of its transient state described above are shown in Table 2.

pc-CRACK software using ASME Code,Section XI elliptical surface crack in infinite plate model and the transient load cases described above as input calculated the fatigue crack growth for the next two operating cycles. Supporting calculations are contained in the project files.

5 RESULTS OF ANALYSIS The observed indication was 66% of the wall thickness at the time of the 1994 UT examination and the ASME Code allowable flaw depth is 75% of the wall thickness. The crack depth, including fatigue crack growth for the period from 1994 to 2005 is 0.7428 inches. The crack depth, including fatigue crack growth, for the next two operating cycles, through April 2010 is 0.7429 inches. This indicates that there is 9% of wall available to accoimnodate any potential future crack growth after two niore operating cycles.

The growth of the indication length is on the same order as the crack depth.

As discussed above, the indication is believed to be associated with original fabrication and is likely subsurface. It is therefore not a serviced induced indication. Potential crack growth mechanisms discussed above indicates that the potential for crack growth is only due to fatigue. SI recommends that UT inspection of this weld be performed at the next outage or the next possible opportunity to characterize the indication more fully. If the indication is shown not to be surface connected to the outside surface, significant additional time could likely be demonstrated.

6 CONCLUSIONS AND 1)ISCUSSIONS Since the end-of-evaluation period flaw depth is well above that for the actual end-of-evaluation period indication, the required ASME Code,Section XI safety factors (2.77 for normal and upset, and 1.39 for emergency and faulted) are maintained throughout at least the next two operating cycles. Based on these fj'Structural Integrity File No.: :CCNIP-06Q-30 1 Revision: 0 42 Associates, Inc. Page6ofl1

results, it is concluded that operation for at least the next two operating cycles is justified with the results, it is concluded that operation for at least the next two operating cycles is justified with the observed indication left as-is.

Table 1: Plant Transient Condition 17; Plant Transient Condition 40 Years Cycle Count Heatup 500 Cooldown 500 Loading 15,000 Unloading 15,000 Step Load Increase 2000 Step Load decrease 2000 Reactor Trip 400 Hydrostatic Test 10 Leak Test 320 Normal Plant Variation 1000000 Loss of Reactor Coolant System 40 Loss of Turbine Generator 40 Loss of Secondary Pressure 5 Table 2: Bendiug Stresses due to Pressure aiid Gravity*

Plant Transient Condition Load Case ID Scale Factor at Beginning Scale Factor at End of Reference Transient per of the Transient the Transient Section 4.2 Heat-up DW 1.000 1.0 Transient 1 Pressure 1.000 NA Transient 1 Steady State 0.8876 N!A Transient 1 Cooldown DW 1.000 1.0 Transient 1 Pressure 1.000 NIA Transient 1 Steady State 0.8876 NIA Transient 1 Loading DW 1.000 1.000 Translent 1 Pressure 0.666 1.044 Transient 1 Steady State 0.8876 1.000 TransIent 1 Unloading DW 1.000 1.000 Transient 1 Pressure 1.013 1.000 Transient 1 Steady State 1.000 0.8876 Transient 1 Step Load Increase OW 1.000 1.000 TransIent 1 Pressure 1.000 1.042 Transient 1 Steady State 0.9831 1.000 Transient 1 Step Load decrease DW 1.000 1.000 Transient 1 Pressure 1.000 0.977 Transient 1 Steady State 1.000 0.9850 Transient 1 StructuralIntegrity File No.:: CCNP-06Q-301 I Revision: 0

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ir I Table 2 (Continued)

Plant Transient Condition Load Case ID Scale Factor at Beginning Scale Factor at End of Reference Transient per of the Transient the Transient Section 4.2 Reactor Trip DW 1.000 1.000 Transient I Pressure 1.000 0.800 Transient I Steady State 1.000 1.000 Transient 1

. Trip NIA 1.000 Transient 3 Hydrostatic Test DW 1.000 1.000 Transient 1 Pressure N/A 1.389 Transient I Steady State 0.617 0.617 Transient I Lealk Test DW 1.000 1.000 Transient 1 Pressure N/A 1.000 Transient 1 Steady State 0.050 0.617 Transient 1 Normal Plant Variation DW 1.000 1.000 Transient I Pressure 1.0DO 1.089 Transient 1 Steady State 1.000 1.000 Transient 1 Loss of Reactor Coolant OW 1.000 1.000 Transient 1 System Pressure 1.0i0 0.7640 TransIent I Steady State 1.000 1.000 Transient I Trip N/A 1.0 Transient 3 Loss of Turbine Generator DW 1.000 1.000 Transient 1 Pressure 1.070 0.7640 Transient 1 Steady State 1.000 1.000 Transient 1 Trip N/A 1.0 Transient 3 Loss of Secondary Pressure DW 1.000 1.000 Transient 1 Pressure 1.000 0.090 Transient 1 Steady State 0.850 0.852 TransIent I LOP N/A 1.000 Transient 2

  • Scale factor detenmined using transient pressure and tetmperature compared to one of the b)ase transients.

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Figure 1: Fatigue Crack Growth Crack Growth 07435 TC U

403000 6oo Bos 1e+0C05 1.2e+005 0 2XoX Cydes Structural Integrity File No.: : CCNP-06Q-30] Revision: 0 C Associates, Inc. Page 9 of 10

I 7 REFERENCES

1. Email fiom Andrew L. Henni (Constellation Energy) to Moses Taylor (SI), dated November 10, 2005;

Subject:

"Shutdown Cooling SS-SS weld inclusion," SI File Number CCNP-06Q-201.

2. E-mail from Tim Lupold (Constellation Energy) to Moses Taylor (SI), dated November 11, 2005;

Subject:

"RE: Calvert Cliffs Shutdown Cooling Outlet Nozzle Safe End-to-Pipe Weld Flaw Evaluation," SI File Number CCNP-06Q-205.

3. ASME Code,Section XI, 1998 Edition.
4. Calvert Cliffs Specification Number 6750-M-3 1A, Revision 2, "Design Specification for Piping, Valves, and Associated Equipment of the Shutdown Cooling System for Calvert Cliffs Nuclear Power Plant Units I and 2," SI File Number CCNP-04Q-203.
5. Calvert Cliffs Report Number 0416750-01, dated September, 1973 (excerpts only), "Calvert Cliffs Nuclear Power Plant Unit 1, Report on the ANSI B31.7 Stress Analysis for Shutdown Cooling Piping System," SI File Number CCNP-06Q-202.
6. ASME Code,Section II, Part D, Material Properties, 1998 Edition.
7. Calvert Cliffs Design Specification Number 8067-31-5, Revision 18, "Project Specification for a Reactor Coolant Pipe and Fittings for Calvert Cliffs 1&2," SI File Number CCNP-06Q-204.
8. pc-CRACKx for Windows, Version 3.1-98348, Structural Integrity Associates, 1998.
9. PIPE-TS2, Version 1.01, Structural Integrity Associates.

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