NL-15-1032, E. I. Hatch, Unit 2 - Inservice Inspection Program Owner'S Activity Report for Outage 2R23
| ML15163A292 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/12/2015 |
| From: | Pierce C Southern Co, Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-15-1032 | |
| Download: ML15163A292 (35) | |
Text
Charles R. Pierce Southern Nuclear Regulatory Affa1rs Director Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham , AL 35242 Tel Fax 205.992 7872 205.992.7601 SOU HERN A .
COMPANY June 12, 2015 Docket Nos.: 50-366 NL-15-1032 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant- Unit 2 lnservice Inspection Program Owner's Activity Report for Outage 2R23 Ladies and Gentlemen:
Enclosed is the ASME Section XI Code Case N-532-4 OAR-1 Owner's Activity Report for the 2R23 Refueling Outage. Table 1, "Items with Flaws or Relevant Conditions that Required Evaluation for Continued Service," lists evaluations performed for continued service. Repair/Replacement activities did occur during Cycle 23 and during the 2R23 Refueling Outage and are addressed in Table 2, "Abstract of Repairs, Replacement or Corrective Measures Required for Continued Service." Per ASME Section XI, Subarticle IWB-31 00, the evaluation of the shroud crack-like indications is provided in Enclosure 4. Please note that information deemed proprietary to EPRI has been redacted, but can be provided to the Nuclear Regulatory Commission upon request.
This report is for the second period of the 41h Interval lSI activities (Interval 4, Period 3, Outage 2).
This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at {205) 992-7369.
Respectfully-"mitted, t.f1.r~
C. R. Pierce Regulatory Affairs Director CRP/RMJ/Iac
Enclosures:
- 1. 2R23 Form OAR-1 Owner's Activity Report
- 2. 2R23 Form OAR-1 Owner's Activity Report, Table 1, Items with Flaws or Relevant Conditions that Required Evaluation for Continued Service
- 3. 2R23 Form OAR-1 Owner's Activity Report, Table 2, Abstract of Repairs, Replacement or Corrective Measures Required for Continued Service
- 4. SNC and Sl Evaluation Core Shroud Axially Oriented Flaw
U.S. Nuclear Regulatory Commission NL-15-1032 Page2 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. A. Vineyard, Vice President- Hatch Mr. M. D. Meier, Vice President- Regulatory Affairs Mr. D. A. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President- Engineering Mr. G. L. Johnson, Regulatory Affairs Manager- Hatch RTYPE: CHA02.004 U.S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. A. E. Martin, NRR Senior Project Manager- Hatch Mr. D. H. Hardage, Senior Resident Inspector- Hatch
Edwin I. Hatch Nuclear Plant- Unit 2 lnservice Inspection Program Owner's Activity Report for Outage 2R23 Enclosure 1 2R23 Form OAR-1 Owner's Activity Report
CASE fcontinued)
CASES OF ,\SME BOILER ,\NO PRESSURE VESSEL CODE N-532-4 FORM OAR*1 OWNER'S ACTIVITY REPORT Repo~Number 2-4-3-2 (Unit 2, 4th Interval, 3rd Period, 2nd Report)
Plant ------------------E_d_"_ l_l_lo_tc_h_N_u_c_lc_o_r_P_In_n_t._P_._o_._B_o_x_2_0_IO_._B_a_x_lc~y-,_G_~_or.s~in_._3_15__
- in__ 13_________________
2 Unit No.--------------- Commercial aervice dote -------'09.;.../0-'5_n_9______ Refueling outage no. _ ___:2:;.;R2;..;;;;;;.,:;.;3_ ___
4th Current inapection intet'lal - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
Uu. lrwl, Jut *lh, oJNrl Jrd Current inspection poriod - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
t11L lncl, lrdl 2001 Edition with 2003 Addenda Edition and Addenda of Section XI applicable to tha inspec1ion plans - - - - - - - - - - - - - - - - - - - - - - - - -
Volume I 12117114 Re,*. S.O; Volume 2 12122114 Rev 4.0, Volum~ 3 1!9i l5 Rev. 4 0 ,
Dote and revision of inspection p l a n s - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
Volume 4: 1/12/15 Rev 6.0, VolumeS 12/18/14 Rt.'V 50, Volum~ 6 12111114 Rev 7 0 Edition end Addenda of Section XI applicable to repoirtreplacament activities. it diftarent than the inspection plans Same N-460, N-663, N-532-4, N-5 13*3, N-586*1 Code Cases u s e d = - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
,., .oD'Iclbfel CERTIFICATE 01' CONFORMANCE Icenily thllt Ia) the statements made In this repon ere corract; tblthe examinations and lists meet the Inspection Plan 11 requored by the ASME Code,Section XI; 1nd lclthe repai 2R23
' /9 ("U>t~
CERTIFICATE OF INSERVICE INSPECTION I, the undersigned. holdino a valid commission issued by the Notional 801rd ol Boiler and Pressure Vasst~llnspectofl and the State or Province of Gegro!a and employed by HSB Global Standards of Hartlort. CT have Inspected the items described in thos Owner's Ac:tlvitv Report, and elate that, to the best of my knowledge and belief, the Owner has performed all activities represented by this repon in accordance with the requirements of Section XI.
By signing this canificate neither the Inspector nor his am player makes any werrantv. t>tprtsaed or implied, concerning tha repair/
replacement activities and evaluation described In th's report. Furthermore, neittler the Inspector nor his omplover shall be liable in any manner for any paraona injUry or propeny damage or a toa of any kind arlainQ from or connected with this inspection.
Date .....~k~-.::3;...*_/.;;.r 3 (N-512*41 SUPP. 9-NC
Edwin I. Hatch Nuclear Plant - Unit 2 lnservice Inspection Program Owner's Activity Report for Outage 2R23 Enclosure 2 2R23 Form OAR-1 Owner's Activity Report, Table 1, Items with Flaws or Relevant Conditions that Required Evaluation for Continued Service
Examination Category and Item Item Description Evaluation Description Number 8-N-2 (813.40) During IWI (VT-3 and EVT-1), crack like indications Evaluated as acceptable for continued service using attached evaluation were reported at various locations on the Core 1500270.301NP.RO and Section XI Shroud. supplement.
D-8(D2.10) After shutdown for the 2R23 outage, leakage was Evaluated per N-513-3 for continued operation. Leak was repaired later in the identified on 8 PSW line feeding the Unit 2 "C" same outage, and is further described on EDG. Table 2 of this report.
F-A(F1.30) RHR Piping Supports (5) failed post maintenance Evaluated as acceptable per RERs SNC585423 and SNC592495 produced by VT-3 due to discrepancies between drawing Hatch Design Engineering.
tolerances and those found in the field.
Monday 19 May 2015 Page 1 of1
Edwin I. Hatch Nuclear Plant- Unit 2 lnservice Inspection Program Owner's Activity Report for Outage 2R23 Enclosure 3 2R23 Form OAR-1 Owner's Activity Report, Table 2, Abstract of Repairs, Replacement or Corrective Measures Required for Continued Service
Code Date Repair/Replacement Plan Class Item Description Description of Work Completed Number 1 "A" FW Check Valve Replaced parts within the FOlOA FW Check Valve after failed LLRT. 3/2/2015 WO SNC637638 Flapped welds 2821-1MS-3-8 and 2821-1MS-3-20 on the Main Steam 1 3" Piping Welds system to identify possible material porosity after lSI indications were 3/5/2015 WO SNC638234 reported.
Full Structural Weld Overlay Installed on Feedwater weld 2821-1FW-12AA-12" FWWeld 1 8. This weld overlay was required due to inspection coverage 3/8/2015 WO SNCS45024 Overlay requirements.
8" PSW Piping to 2C Replaced section of 8" PSW line that was leaking due to a small hole. This 3 3/25/2015 WO SNC636435 EDG piping feeds the Unit 2 "C" Emergency Diesel Generator.
RHR and Core Spray room cooler PSW return line was isolated and repaired 3 3" PSW line 4/27/2014 WO SNC569457 after leaking pipe near 2P41FOOGA was identified.
Wednesday 19 May, 2015 Page 1of2
3 2" PSW line During an lSI pressure test for the PSW system, a leak was identified on the 1/16/2014 WO SNC533953 discharge piping coming from the "A" Control Rod Drive room cooler. The leak immediately upstream of the 2P41F002A valve was isolated and repaired.
Wednesday 19 May, 2015 Page2of2
Edwin I. Hatch Nuclear Plant - Unit 2 lnservice Inspection Program Owner's Activity Report for Outage 2R23 Enclosure 4 SNC and Sl Evaluation Core Shroud Axially Oriented Flaw
Soutbtm Nuclear c Wartr NMP-E8-CJ18.(X]B Instruction lndlcdan Natlflcdcn Vfilllan 3.0 1af2 i i 1 5 l;&lhiiiDitlnC PTO MTO ETC RTD VfXOiherD ~~U.1 lladzlalVVI Dllba Con l!lraaf. auk lib ladlclda* wnrepodal It 1111 tbUowfq laclriaa:L Selllllldlld JNJt'* tlr llldlllaall dabdls. Bd:lrau:o all0031'730.
2Blll\IIJ3. SlrDad Vlr1lcll Wild .J20- Wtkl V-4 haiiD (EYT-l).INR H2R23-IVVI-J.s.o7 2BU 1\1115, Sllraud V.:tiad Wllld%WWIId V-6 hal m (EVI'*l).INR.J1212HVVJ.J5-0f D111\BJJO.Slmld:laiBriarSartllcl:llamTap06Rupsat1Uq.n--4,-(Caii42-47)(VT"Il)JNllBZR23-I'M-1S-03.
21U 1\BW, Jllnudlallrfor..._ hill Tap Qaldl alpplrtJUiri. 45"*"* (CIIIl44-43) CVT*3) "-HD23-IVVI-15-06.
DU 1\BlSO. Sllnlad J.alldor 8artlcts tam Tap Ollldll Jas-..145* (CeD 42-D7) (VT*3) lNilll2123-lVVI*tS*05, (l}bliiw.-:
Pleaao aee aetached BWRVIP and Saceian XI avaluaeions far further detail*.
MO corrective action required, please aee attached evaluations far detaila.
[
- I hri**' 7 I iljj;i i§iiilid zhiil F1ple 1
Southern Nuclear Operating Company Nuclear NMP*ES-01 0*200-FO 1
~ Management Hatch Vessel & Internals Program - Flaw Version 1.0 EvaluaHon Form
~"'""*'- ......... Form Page 1 of 3 BWRVIP EVALUAnON TE II CAR25S483 Component Core Shroud Applicable NMG NMP-ES-010*GL02 BWAVIP GuldeUne I# BWAVIP-76. Rev. 1 INF/INA # 115H2005 Condition:
During 2A23 Hatch pertorroed oJqnlflcanl shroud visual egms frpm both lbe ln!erJO! and ex1er1or shroud surfaces. Cwe Shroud interior surfaces at cells C42-47). 146-431. (42-07). 106-111 and t10-4n were exams from b9th the 00 and 10 surf8ces. Fye OEH IN As dgcumenttha findings frgm theg shrouc!
insoections:
- 1. INA 1§..03
- VT-3 exam of con lane (42-471 id9nt!lied five tgtal axially orlenled flaws measuring a maximum of 1.ss*. One of these flaws crosses the entire H-4 weld.
- 2. INB 15*05- Shroud lane 142=07) identi!ied four total flaws in the ylciollV of the V-3/H-4 Intersection. !he largest of which was 6.3" runnlna §Omawbat parre!lello the Y-3 weld and cass!na comR!ete!y tbrgyqb th H-4 weld.
- 3. INA 1§..06 *Two Rawest arqas were lc!entUied durtoo eqm!natjons gf cell lane 146-43). bo!h or Which were outside of weld heat affected zones. Two indications measuring a maxtmum of 1.9..
were found rouahly a* ccw from V*5 and 12" Inches aboye H-5. Thege ind"JCations are uiafty oriented and located In an area showing signs of heavv surface grlndlna. Two additional axlaUv oriented Indications to.aa~ maxjmuml were found lust below !he H-5 weld and s!ighllv CW of V-5.
The flaws described In lbe above 3 !NRs were initially fden!lliecl yia the MIME section XI VI-3 exam tad UPOn slscoyerv were lyrlt!er eyaluamd wi!h Jht eyr-1 method. The llaws listed In the 2 lfl!As below were lqen!jfied yla BWRV!P oresc;r!bed EVT*l &!!§tnS*
- 4. INR 15-04
- M*l exarna oerfonned on vertical weld V*6 ogt§Si two axiallv oJlented (ls,fica!lgns wjlh a my!mum lenglb of 3.1". One of these Indications starts on the ccw side gf the V-6 wa~
toe and extends URWPrd. through the H-4 weld and beYOnd the beat affeqtecl zone abQye H*4. It should be noted that lhe previously noted V6 indication tbal was tracked during 1818 and reinspected 2619 was d8teqn!ned to be non-relevant d!Jring 2623. although the non-relevant
- ~
Nuclear Management Form Southern Nuclear Operating Company Hatch Vessel & Internals Program- Raw Evaluation Form NMP-ES-01 0-200-F01 Version 1.0 Paae 2 of 3 Indication seemed lg and where Jhe 2R23!nQjs;atlcm begins at the CCW watd lo of V*6. The OI!Jer 2623 Indication alibis lgcaUQO branches gff of ths fjret3.1" lndicatjon lust above V-6 on the uDDer edge of H-4 and ruosaany aWJy frprn H-4 for rqudJ!y 1.1".
- 5. INA 15-DZ- M-1 egm!noUgo of wsld V-4 DQied a 3.5" axially odemacllnd!cotion running fmm ab9ve the H-4 weld rgyQh!v 1.s* cw of V-4 tg a locatlgn myqhlv0.92" below H-4.
Disposition:
SIOJCturaflntegrllv A5aoclotes was contacted during the 2R23 outage to help Halcb evaluate lbt referenced condillon. S!A ca!culltion 1500270.3Q1-RO constfVIIlvetv usumes a!lldent!Oed flaws " '
!brough-wa!l !n d!Pih. 11tbtn eya!uatu a tlngJe botJndjng flaw (6.3" flaw at \1-3/H-4), haying sbawn that parallel naws are bounded bY a s!og!g flaw. Ib!* avaluation calculatot the maximum OUg,wabla flaw prior to 2R26 SladwJ. !law eyaluatioN II5C!Claled with bJte \YQid§ must be Allbee rHyi!UO!ecf assumjoq 111 weld ligament above this flueoce is flawed. or the existing eyatuatlon subm!lt!!d to the NBC.
Basis for Dlsposllion:
All axlallv oriented jndlca!lons idlntified dUJinq 2823 are boJ.Indod by me 8.3" flaw which was delermin!d acceptable In SIA calcu!atiOO 1500270.301-RO. While INA H2823.JWI*15*03 and 15::05 dg lden!ilv borlzo(l!allndiCJtl!iKlJ alona lha H-4 weld. !be tie rocfs installed on the shroud II[Jieturallv replace !he horizcnlal welds which mitigate horizontally orjeoted flaws. !hi above referenced calculation estab!jshes Scope Expansion (ref. NMI step 4.6.9)
I
- no scooe expansion I& regy!rtsl as all yertical weld$ have established Intervals for rt-IO&QiCCtion. Hpr!zontal walcfs to not reqUire examination due to Installed Ue rods* See allat;hed document for lustiflcallan aqalns! ASME Section XI scope e!!DJOIIgn.
Southern Nuclear_QIJ'!ratlng Company Nuclear NMP-ES*010-200*F01
~
Hatch Vessel & lntemals Program - Flaw Management Version 1.0 Evaluation Form
.......- .... a.~, Form Paae3 of3 Follow-on or Supplemental Examinations (ref. NMI step 4.6.1 0)
The above-referenced structural evaluation justifies a 10-year re*lnspecllon intervaL however due lo actgiJ!oolll eva/ygtion I!DJitatlpns which wnt be apgUcab!e after 2R2§. re=examioalion of \1-4. v-s. and flaws reoortecl at litnes 42-47. 46-43<V-5l. and 42-07lV-3l ace recommended for relnspection during 2R26 to apply up to di!te jospectlon results for re-evaluation. In addiUon. the flaws seen In ceUs 46-43.
42.()7 and 42:47 thai are not usoclatecl wtth ellher vertical or hgrizontal welds will be dealt with usJng the rules wl!hjn ASME Section XI. Because or these ASME code rules. shroud 10 lanes !42-471. (42.P7l and
!46-43) mus! be examined using the VT-3 method in each or the three lSI periods In the 5111 1SIIntewl beginning 1 January 2016. See aftached d9cumenl for AaME Section XI re-eurnlnaUon rreaueoc;v discussion.
Rererences (ref. NM! step 4.7. 1)
INA H2A23-IW1*15-03 Core Shroud 10 from 42-47 INA H2B23*1YVH5*04 Core Shroud Vertical Weld V-6 INA H2R23-IWI*15*05 Core Shroud tp from 42-97 INA H2A23-IYYJ-1S-OBCore 5hroud 10 from 46*43 INA H2R23*1WI*15=07 Core Sllroud Vertical Weld V-4 SIA documenl1500370.301 Rev. 0. HNP? Core Shroud Axlp!ly Oriented Flaw Calculation Review and Approval (ref. NMI Step 4.7)
Responsible Engineer: Andrew Gordon I d.J.J./-1-. oate: ,y,vz~
Independent Review: Deliaa Pouroaras tM,, J fiw. ,_.........Date: W7,/~t2" Supervisor Approval: P.,f, V,J4...1 D~/1!. W.1wrJ
Edwin I. Hatch Nuclear Plant- Unit 2 Technical Justification Regarding S<:ope Expansion of ASME Code Section XI B*N-2 Components On February 21,2015, Inspection personnel performing scheduled visual VT*3 examinations of the Edwin I. Hatch Nuclear Plant- Unil 2 Core Shroud reported new crack-like Indications not associated with a horizontal or vertical weld. The Indications were repor1ed to SNC via INA H2R23-IWI*15*03,*05 and -06 and are located at 40e, 45*
and 140° azimuths and In the areas of and between horizontal welds H4 and HS.
The two indlcallons detected at 400 from cell location 46-43 were measured to be 1.7" and 1.9" in length. These indications were located rou9hly a* ccw from the V-5 weld and 12* above the H-S weld. As noted In INA 15-06, Inspection personnel noted evidence of heavy surface grinding In the area.
An addltlonallndicatlon extending beyond lhe H-4 HAZ was identified at 45* during inspection of cell lane 42-47. Horlzontallnc:llcallons were noted within the weld HAZ.
although one Indication branches axially outside of the HAZ. This Indication branches up and away from H*4 extending a distance of 1.65" beyond the toe of H-4. Four other Indications were noted within INA 15-03 but in all Instances these other indications are associated with the H-4 weld.
The final indications are noted within INR 15-05 at roughly 140° near the V-3/ H-4 Intersection seen In cell 42.07. The major Indication noted here Is 6.3" In length which runs roughly parallel to V-3. This Indication starts roughly 1" CW of V-31n the base melal and terminates below H-4, still approximately 0.758 CW of V-3. Two additional axial indications measuring 0.9" and 1.4" are seen below the H-4 weld but outsida of Its HAZ. A fourth indication of 1.3" In length is seen branching axially below the toe of H-4.
These examinations were performed to comply with ASME Section XI Code iable IWB*
2S00-1, category B-N-2 Item Number 813.40 "Core Support Structures" requirements and therefore the rules of IWB-2420 usuccessiVe Inspections" and IWB-2430 "Additional Examinations* apply. This Technical Justification Regarding Scope Expansion for Examination has thus been prepared to fay the requirements and logic for compliance with these requirements.
Successive Inspections The applicable rule for compliance regarding successive examinations is IWB-242D(b) which requires that "the areas containing thf'J flaws or relevant conditions shaH be 1eexsmined during the nSKt thrss iMpeotlon periods listed In the sChedule of the inspection program of IWB*2400." IWB-2420(c) prascrbls that "'I.. .the flaws or 1elevsnt conditions remain unchanged lhf'J component examination may revert to rhe orlgfnal schedule of successive Inspections." The 2A23 refueling outage Is the last outage of Period 3 of lSI Interval 4. Compliance thus dictates that follow-up examinations be conducted during an outage ln each of the three periods of the s"' Interval.
Additional Examinations IWB-2430(a) states that In part "Examinations performad In accordance with Table IWB-2500-1 that reveal/laws or relevant conditions exceeding the acceptance standards of Table IWB-3410-1 shall be extended to include additional examinations during lha cuff81Jt outage. The additional examinations shall include an additional number of welds, areas, or parts, Included in the
/nspecllon item, equal to the number of welds, areas, or parts, Included In the Inspection Item that were scheduled to be performed during the present inspsclion period. The additional examinations shall be selected from welds, areas, or parts of similar material and service.
- The original Code Section XJ requirement for B-N-2 Item 813.40 is "Accessible Surfaces" employing visual VT-3 examination techniques.
In order to assess polentlallocations for scope expansion the original examination requirement must be assessed. The Inner diameter of the core shroud Is considered a B-N-2 component In Its entirety. Thus all accessible surfaces would require examination during each 10-year lSI Interval and are scheduled accordingly. The shroud surfaces which are considered "always" accessible ara those surfaces which can be examined from the cuter arameter of the shroud because the only obstructions are permanent ones such as jet pumps or Ue rods. These Items are scheduled throughout the Interval by lanes corresponding to azimuths. Unlike the outer diameter, the Inner diameter of the shroud along with any surfaces below the core plata do not nonnally become accessible for VT-3 surface examination. However, scheduled fuel cell evacuations along the periphery of the core permit some azimuths of the Inner diameter to be accessible for VT-3 surface examination and they are scheduled appropriately. The Inner diameter surfaces are also scheduled by azimuthal lanes corresponding to the fuel cells which are evacuated during the outage. The 2R23 outage scope contains those azimuthal lanes that were to be evacuated. In total, five fuel cens were made available fer Inspection, and all were Inspected. It Is not an ASME Code Section XI requirement to evacuate fuel cells or remove other components to "make" surfaces accessible.
Basis for sufficiency- The indications detected were In base metal material not associated with any horizontal or vertical shroud weld. Indications that were noted In exam reports to be associated with welds are handled within appropriate, approved BWRVIP guidance. Axial indications noted within this document were evaluated within Sbucb.lrallntegrity Associates calculalfon 1500270.301 Rev 0. This calculation demonstrates the structural acceptablnty of aU indications for at least 10 years. Those indications noted lo be associated with surfaces and not welds, wiU be scheduled fer successive examination In eac:f'l of the 3 Periods within the s" lSI Interval in accordance with the Section XI code.
Based on this assessment additional visual VT*3 examinations are not warranted to meet Code requirements.
Extent of Condition Extent of condition exams should be discussed by the BWRVIP Technical Team, with Input from lSI personnel. Based on the recommendaUon from this discussion, additional
- surface examinations In base material, particularly in areas that have not been examined during the 10-year lSI Interval may be scheduled. TE 916111 has been Issued to ensure further discussion of this topic at the next opportunity*
~Structural Integrity Associates, Inc. File No.: 1500270.301NP Project No.: 1500270 CALCULATION PACKAGE Quality Program: [gl Nuclear 0 Commercial PROJECT NAME:
HNP2 Shroud Flaw Evaluation CONTRACT NO.:
Purchase Order: SNG 19354-0012, Rev. 0 CLIENT: PLANT:
Southern Nuclear Operating Company, Inc. Edwin I. Hatch Nuclear Plant, Unit 2 CALCULATION TITLE:
Hatch Nuclear Power Plant Unit 2 Core Shroud Axially Oriented Flaw NOTE: This document is the non-proprietary version ofSI Calculation No. 1500270.301 , Rev. 0. The EPRI and Southern Nuclear proprietary information has been redacted.
Project Manager Preparer(s) &
Document Affected Revision Description Approval Cbecker(s)
Revision Pages Signatures & Date Signature & Date 0 1 - 18 Initial Issue Resuonsible Engineer CA!Z ~1-~
r Daniel V. Sommerville Heather F. Jackson 05MAR2015 05MAR2015 Resuonsible Verifier
~/.-~L- . ---
-~~;~:::.:---- - - -
Fabien Hsu 05MAR2015 Page 1 of 18 F0306-0I Rl
e Structural Integrity Associates, lnc..a Table of Contents
1.0 INTRODUCTION
......................................................................................................... 4 2.0 OBJECTIVES ................................................................................................................ 4 3.0 METHODOLOGY ........................................................................................................ 4 3.1 Flaw Evaluation ................................................................................................. 4 3.1.1 Characterize Flaws ............................................................................................ 5 3.1.2 Material Properties ........................................................................................... 5 3.1.3 Inspection Uncertainty....................................................................................... 5 3.1. 4 Crack Growth ................................................................................................. ... 5 3.1. 5 Fracture Mechanics ........................................................................................... 6 4.0 DESIGN INPUTS .......................................................................................................... 8 5.0 ASSUMPTIONS .......................................................................................................... 10 6.0 CALCULATIONS ....................................................................................................... 10 7.0 RESULTS .................................................................................................................... 11
8.0 CONCLUSION
S ......................................................................................................... 11
9.0 REFERENCES
............................................................................................................ 12 File No.: 1500270.301NP Page 2 of 18 Revision: 0 F0306-0IRI
e Structural Integrity Associates, Inc."'
List of Figures Figure I. HNP2 Core Shroud Configuration [II] .................................................................. 15 Figure 2. Map ofHNP2 core shroud indications based on 2R23 inspection results [1].
Annotations show approximate location, orientation, and length in inches. OD indications originally reported at H4 weld at 75° azimuth (gray lines in figure) were subsequently removed by brushing ....................................................................... 16 Figure 3. LEFM Solution for a Single Through-wall Axial Crack in an Internally Pressurized Cylinder [1 0] .......................................................................................................... 17 Figure 4. LEFM Solution for an Infinite Array of Through-wall Cracks in a Plate with a Membrane Load [I 0] ............................................................................................. 18 List of Tables Table I. HNP2 Core Shroud Evaluation Fluence and Failure Modes ..................................... 14 Table 2. HNP2 Shroud Bounding Axial Flaw Evaluation Results .......................................... l4 File No.: 1500270.301NP Page 3 of 18 Revision: 0 F0306*01RI
S} Structural Integrity Associates, Inc.*
1.0 INTRODUCTION
Multiple reportable indications were detected in the Hatch Nuclear Plant, Unit 2 (HNP2) core shroud during the Spring 2015 (2R23) refueling outage planned inspections [1].
The 2R23 inspections [2] consisted of:
- Visual examinations (VT) ofthe inside and outside surfaces of vertical welds V3, V4, V5, V6, V7, and V8,
- Shroud exterior surface VT examinations from top guide to core plate at azimuths of 35 to 65°,
125 to I55°, 2I5 to 245°, and 305 to 335°,
- Shroud interior surface VT examinations from top guide to core plate from azimuths 90 to 270°.
Several axially oriented indications were reported on the inside surface of the core shroud in the vicinity of the core shroud circumferential welds H4 and H5 and in the base metal near the vertical weld V5 intersection with circumferential weld H5 [1].
Figure 1 illustrates the HNP2 core shroud configuration and weld locations, while Figure 2 illustrates the locations and sizes of reported indications, based on 2R23 inspection results [I].
2.0 OBJECTIVES The objectives of the work documented in this calculation package are to:
- 1. Perform a flaw evaluation for all of the axially oriented indications reported during the Spring 2015 HNP2 refueling outage (2R23).
- 2. Perform a flaw evaluation of the base material indications in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)
Code,Section XI rules for inservice inspection of light water reactors.
Circumferentially oriented indications or the circumferential component of indications are not evaluated in this calculation package since the tie rod shroud repair installed at HNP2 structurally replaces the circumferential welds [2]. By extension, a circumferentially oriented flaw elsewhere in the core shroud is not structurally significant since the tie rods provide the load-carrying capacity for lateral loads.
3.0 METHODOLOGY The methodology used for the flaw evaluation and leakage calculation is discussed below.
3.1 Flaw Evaluation Evaluation of the axially oriented flaws is performed using methods consistent with those presented in BWRVIP-76, Revision I [3]. The following process is used:
- 1. Characterize location and dimensions of all reportable indications, using the most recent shroud inspection data,
- 2. Select applicable material properties and failure modes,
- 3. Apply inspection uncertainty as appropriate for the method and delivery system used,
- 4. Add crack growth for the applicable evaluation interval and growth mechanisms, File No.: 1500270.301NP Page 4 of 18 Revision: 0 F0306-0IRI
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- 5. Evaluate stability of the indications using the fracture mechanics methods appropriate for the material type and environmental conditions.
It is important to note that the flaw evaluation methods provided in Reference [3] are based on the rules of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI IWB-3600 and Nonmandatory Appendix C. Material-specific crack growth rates, tensile and fracture toughness properties, and inspection uncertainties, which are not provided in ASME XI, are used in this evaluation since they are applicable to the environment, material, and inspection methods.
All aspects of this methodology have been reviewed and accepted by the U.S. Nuclear Regulatory Commission (NRC) with respect to inspection and evaluation of the Boiling Water Reactor (BWR) core shroud [3].
Each step of the process is described separately below.
3.1.1 Characterize Flaws The number, orientation, and dimensions of the reported indications are obtained from the inspection notification reports (INRs) provided by General Electric Hitachi Nuclear Energy Americas, LLC (GE-H)
[1]. The INRs are also used to infer the likely initiation and growth mechanism for the reportable indications which is necessary in order to identify which crack growth mechanisms to consider in the flaw evaluation.
Several indications are reported on the core shroud inside surface [1]. All of the indications on the shroud inner surface are conservatively treated as through-wall flaws. A single bounding flaw is selected for evaluation rather than individually evaluating each of the indications reported. See Assumptions 1 and 5 below.
3.1.2 Material Properties BWRVIP-100, Revision 1 [4] is used to identify the failure modes appropriate for the level affluence in the core shroud material. Tensile properties will be selected at a temperature and fluence level such that the allowable flaw sizes and plastic zone size are bounding. BWRVIP-100, Revision 1 [4] and the ASME Code,Section II, Part D [Sa] are also used to select the appropriate tensile properties for the fluence level and temperature. BWRVIP-100, Revision 1 [4] is used to select the appropriate fracture toughness for the shroud fluence. The results of this procedure are described in Section 7.0.
The peak tluence on the shroud at the end of life, which is reported to be 50.1 effective full-power years (EFPY) and corresponds to the year 2038 [6] is conservatively used for this evaluation.
3.1.3 Inspection Uncertainty Inspection uncertainties provided in the inspection method demonstration documentation, appropriate to the inspection method and delivery system, are applied to the length of all reportable indications [7].
3.1.4 Crack Growth Consistent with the methods provided in BWRVIP-76, Revision 1 [3] and the clarifying guidance given in References [8, 9] intergranular stress corrosion crack growth (IGSCC) is calculated for the evaluation File No.: 1500270.301NP Page 5 of 18 Revision: 0 F0306-0IRI
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interval and added to each end of each reportable indication. Fatigue crack growth is not a relevant mechanism for the core shroud; therefore, IGSCC is the only relevant crack growth mechanism.
The IGSCC length crack growth rate provided in BWRVIP-76, Revision 1 [3] and BWRVIP-14-A [8] is used for all flaws:
daldt = {I }/
Each tip of each flaw will grow by 4.38 inches during the 10-year interval as shown below:
({( )}j (1 0 years) (365.25 days/year) (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s/day) = 4.38 inches I flaw tip 3.1.5 Fracture Mechanics Stability of the core shroud is assessed by performing both limit load and linear elastic fracture mechanics (LEFM) calculations, based on the material properties established in Section 4.0. The limit load calculations evaluate stability ofthe core shroud structure; whereas, the LEFM calculations evaluate stability of the crack.
The limit load method provided in ASME B&PV Code,Section XI, Nonmandatory Appendix C [5b] for through-wall axial flaws is used since the method provided in BWRVIP-76 Revision 1 [3] assumes that the circumferential welds on either end of the vertical weld are cracked through-wall. For the present situation, some of the flaws to be evaluated are at, or passing through, a circumferential weld. Therefore, the assumption of a finite height cylinder is not well suited. Consequently, the ASME B&PV Code Section XI, Nonmandatory Appendix C solution is considered a more appropriate solution. The following equation is specified in Section XI C-541 0 for through-wall axial flaws [5b]:
2 U flow
/allow = 1.58 * ~ * -- -1 (1)
( UHoop )
Where fallow 2a, allowable flaw length, inch Rm = Shroud mean radius, inch t Shroud wall thickness inch O"tlow = Flow stress, ksi O"Hoop Hoop stress, ksi To remain consistent with the intent of ASME Section XI [5b] and based on guidance in the technical basis documentation [14], a Structural Factor (SF) for the appropriate Service Level condition is applied to hoop stress. Service Level D is the limiting condition, and an SF of 1.39 is used as recommended in BWRVIP-76, Revision 1 [3, Appendix 0.5].
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e Structural Integrity Associates, lnc.c LEFM solutions published in Reference [1 0] are selected for this evaluation, as described below. The solutions used for this evaluation are consistent with those suggested in BWRVIP-76, Revision 1 [3].
The following LEFM solutions are used:
- Single through-wall axial flaw in an internally pressurized cylinder (See Figure 3) [1 0, pg. 485],
- Infinite array of parallel through-wall flaws in a plate subjected to a membrane load (See Figure 4) [ 10, pg. 256],
- Plastic zone size correction [ 10, pg. 16].
The LEFM solution for a single through-wall axial flaw in an internally pressurized cylinder is a good representation of a single axial flaw in the core shroud. The LEFM solution for an infinite array of parallel through-wall flaws in a plate provides means of understanding the interaction between multiple parallel flaws. This solution is used to show that treating a single flaw by itself provides a bounding treatment of the driving force at the tip of the axial flaw. In other words, review of the LEFM solution for an array of parallel axial flaws shows that adjacent flaws tend to "shield" each other and reduce the resulting driving force at the crack tip.
The radius of the plastic zone size is added as an additional crack length at each end of each flaw. The plastic zone size correction is estimated, for conditions of small scale yielding, using the following equation [10, pg. 16]:
(2)
Where is used to adjust for plane strain or plane stress conditions at the crack tip, where:
Plane Strain: a =1/67t Plane Stress: a =1/27t is yield stress, ksi For this calculation the allowable fracture toughness, K1c, and the plane stress adjustment is used with Eq. (2) above to obtain a bounding estimate of the plastic zone size for the flaw stability calculations, regardless of end of interval flaw size.
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e Structural Integrity Associates, lnc.ft 4.0 DESIGN INPUTS The design inputs used for this calculation are identified below:
Geometry:
The shroud geometry is taken from Reference [II]:
- Shroud ID: 174.5 inch [II a, for elevations between top guide and core plate]
- Shroud Thickness, t: 1.5 inch [ 11 b]
Loads and Through-wall Stress Distributions:
The upper shroud RIPD values are taken from Reference [2], and are summarized as follows:
- Level A RIPD: 7.8I psi
- Level B RIPD: 11.72 psi
- Level C RIPD: 29.5 psi
- Level D RIPD: 29.5 psi IGSCC Crack Growth Rate:
The IGSCC length crack growth rate provided in BWRVIP-76, Revision I [3, page F-I] and BWRVIP-I4-A [8, page 6-2) is assumed at each crack tip.
This crack growth rate in the axial direction is (( ll [3, 8].
Reactor Coolant Water Chemistry:
HNP2 implemented hydrogen water chemistry (HWC) in September 1991, Noble Metal Chemical Addition (NMCA) in March 2000, and began On-Line NobleChem (OLNC) in November 20II [2, 12].
Under HWC conditions and OLNC, the shroud horizontal welds H3, H4 and H5 are considered mitigated, and the vertical welds between these horizontal welds (shown in Figure I and Figure 2) are also considered mitigated [ 13, Table 4-I]. However, crack growth based on normal water chemistry is conservatively assumed in this calculation. See Assumption 6 below.
Shroud Fluence:
The peak shroud tluence along the entire core shroud height at the end of design life at 50.1 EFPY (2038) is conservatively used for this evaluation, per Assumption 2 below. The bounding fluence at 50.1 EFPY is (( [6]. Material Type: The shroud material is SA-240 TP304L stainless steel [lla]. File No.: 1500270.301NP Page 8 of 18 Revision: 0 F0306-0JRJ
e Structural Integrity Associates, lnc.c Material Properties: For fluence values greater than (( }/ but lower than I { } }, LEFM analysis should be used to determine the structural stability of flaws in the core shroud, based on a static initiation, plane strain, mode I fracture toughness. Consequently, the following toughness value is used for this evaluation:
- II II )} [4]
The material flow stress and yield stress both increase with fluence [4]. However, it is conservative to use un-irradiated materials properties since this will result in a larger plastic zone size and a smaller allowable flaw size. Consequently, un-irradiated tensile properties are used [5a]:
- cru (un-irradiated, 550°F): 57.2 ksi [5a Table U]
- cry (un-irradiated, 550°F): 15.9 ksi [5a, Table Y-1]
- crr (un-irradiated, 550°F): 36.6 ksi (taken as the average of cru and cry)
Initial Flaw Distribution: The flaw lengths and configurations are taken from the 2015 INRs [1]. Since several indications are reported, and many have been detected using a VT-3 inspection technique (accuracy of sizing is uncertain), no attempt is made to evaluate every indication in this calculation package; rather a bounding approach to flaw evaluation is taken as discussed in the methodology section. Inspection Uncertainty: Evaluation factors to account for inspection uncertainties for the visual (EVT-1 and VT-3) inspection data are taken from the applicable demonstrations for the inspection technique identified in the INRs [1]. The evaluation factors are taken from BWRVIP-03 [7, Section 3.1]. No depth evaluation factors are used since all flaws are treated as through-wall. The applicable length evaluation factor associated with VT and measurement of flaws with a ruler is (( }} [7, Section 3. I]. This factor is applied at each axial crack front of the evaluated configuration, per Assumption 7 below. Operating Cycle Duration: HNP2 is on a 2 year operating cycle [2]. File No.: 1500270.301NP Page 9 of 18 Revision: 0 F0306-0I Rl
e Structural Integrity Associates, Inc.* 5.0 ASSUMPTIONS The following assumptions are used in this evaluation.
- 1. All flaws are assumed to be through-wall for the structural evaluation.
This assumption is bounding and necessary since the inspection technique reported in Reference [1] is visual inspection (VT) and is not capable of detectingflaw depth. This assumption provides a boundingflaw evaluation for a// flaws.
- 2. The 50.1 EFPY peak shroud fluence is used to determine the fracture toughness for all flaws.
This assumption is conservative because it applies the boundingfluence projected at the end of design life to a// locations. The current refueling outage is 2R23 and the end of design life is reported to be 50.1 EFPY, which corresponds to the year 2038 [6].
- 3. A 100% capacity factor is assumed for crack growth.
This assumption is conservative because it uses the maximum number of hours possible, each year, for crack growth.
- 4. Flaws are allowed to grow through the horizontal welds.
Inspection data .from the HNP2 core shroud shows evidence offlaws growing through the horizontal weld H4 {1]; therefore, this assumption is considered appropriate.
- 5. A single bounding flaw is evaluated in this calculation package which is defined to bound the length reported for all axial indications.
Rather than evaluate all axially-oriented indications separately, a single flaw evaluation is performed of the single largest flaw. It is shown in this calculation that parallel flaws are bounded by a single flaw; thus, this approach bounds all reported flaw lengths and configurations (single or multiple parallel flaws).
- 6. Normal water chemistry is used.
The use of normal water chemistry is conservative since it assumes the fastest crack growth rate.
- 7. A length evaluation factor of (( }} is used when evaluating flaws.
This assumption is conservative since no adjustment to measured flaw length is required for the purpose offlaw evaluation. This assumption is applicable to flaw lengths obtained using VT and measurement with a ruler, and when the RMS value offlaw length measurement errors during performance demonstration is less than 0.75 inch, as cited in BWRVIP-03, Revision 17, based on documentation in BWRVIP letter 2004-426 {7, note 3 to Table 3.1-1]. 6.0 CALCULATIONS All calculations are performed using an Excel spreadsheet, and flaw stability is assessed by performing bounding calculations for all reported indications, as follows:
- 1. The bounding shroud fluence at 50.1 EFPY is applied for all shroud elevations resulting in a lower-bound fracture toughness for all flawed locations.
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- 2. Review of the LEFM solution shown in Figure 4 illustrates that the crack driving force for parallel flaws in a plate is reduced compared to the case of a single flaw in a plate, as shown by observation of the Fu curve for s---+0 which corresponds to a single flaw. Consequently, multiple aligned flaws can be bounded by treating each as a separate flaw.
- 3. The longest flaw reported near the H4/V3 intersection is selected for evaluation and clearly bounds all flaw lengths reported. This flaw is assumed to be a single continuous flaw passing through the H4 weld. The initial flaw size considered, before addition of uncertainty and crack growth, is 6.3 inches [1c].
7.0RESULTS Table 1 summarizes the applicable failure mode and analysis method based on the end-of-interval fluence. The allowable flaw size for an axially oriented flaw in the HNP2 core shroud, considering the lower-bound fracture toughness of 1.{ /1. and un-irradiated yield strength (conservative) is: LEFM: 46.02 inches Limit Load: 275.40 inches Table 2 presents the results of the axial flaw LEFM evaluation. The 10-year, end of interval, bounding flaw size is 15.5 inches. This flaw bounds all reported axially oriented indications. The final flaw size is determined by adding to the reported flaw size the inspection uncertainty and projected 10-year crack growth at each crack tip: End-of-interval flaw size= 6.3 inches+ 2(0.20 inch)+ 2(4.38 inches)= 15.5 inches The calculated allowable operating interval for the reported indications is approximately 44 years using the bounding IGSCC crack growth rate. No re-inspection interval greater than 10 years is currently allowed in BWRVIP-76 [3]. After 10 years of operation from the Spring 2015 inspections, the core shroud structural margin, on fracture toughness, is 3.96 (Table 2). The required structural margin is 1.39.
8.0 CONCLUSION
S The results of flaw evaluations show that the axially oriented indications reported in the HNP2 core shroud during the 2R23 outage are acceptable as-is for a 10-year inspection interval. The calculated inspection interval for all axially oriented flaws is greater than 10 years; however, the required re-inspection interval is as defined by the applicable re-inspection requirements provided in the ASME B&PV Code Sec. XI (Code year and addenda as approved for the current operating interval for the plant) [5] or BWRVIP-76, Revision 1 [3], as appropriate, depending on whether the flaws are located in the base material or adjacent to the core shroud welds. In any case, flaw evaluation results support a 10-year inspection interval. File No.: 1500270.301NP Page 11 of 18 Revision: 0 F0306-0IRI
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9.0 REFERENCES
- 1. H2R23 Inspection Results, GE-H Indication Notification Reports, dated February 23, 2015, SI File No. 1500270.202:
- a. INR H2R23 IVVI-15 Shroud 10 Surface from Top Guide Cell42-47
- b. INR H2R23 IVVI-15 Shroud 10 Vertical Weld V06
- c. INR H2R23 IVVI-15 Shroud 10 Surface from Top Guide Cell 42-07
- d. INR H2R23 IVVI-15 Shroud ID Surface from Top Guide Cell 46-43
- e. INR H2R23 IVVI-15 Shroud 10 Vertical Weld V04
- 2. Design Input Request, Revision ] ,received February 27,2015, SI File No. 1500270.201.
- 3. BWRVIP-76, Revision 1: BWR Vessel and Internals Project, BWR Core Shroud Inspection and Flaw Evaluation Guidelines. EPRI, Palo Alto, CA: 2011. 1022843. EPRI PROPRIETARY INFORMATION.
- 4. BWRVIP-100, Revision 1: BWR Vessel and Internals Project, Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds. EPRI, Palo Alto, CA:
2010. 1021001. EPRI PROPRIETARY INFORMATION.
- 5. ASME Boiler & Pressure Vessel Code:
- a. Section II Part D, 2001 Edition through 2003 Addenda.
- b. Section XI, 200 I Edition through 2003 Addenda.
- 6. Trans Ware Report No. SNC-HA2-001-R-OOI, Revision 0, "Edwin I. Hatch Unit 2 Fluence Evaluation at End of Cycle 22 and 50.1 EFPY," April 2014, SI File No. 1500270.205P.
SOUTHERN NUCLEAR PROPRIETARY INFORMATION.
- 7. TR-105696-R17 (BWRVIP-03) Revision 17: BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines. EPRI, Palo Alto, CA: 2014. 3002003091. EPRI PROPRIETARY INFORMATION.
- 8. BWRVIP-14-A: BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Stainless Steel RPV Internals. EPRI, Palo Alto, CA: 2008. 1016569. EPRI PROPRIETARY INFORMATION.
- 9. BWRVIP Letter 2012-074 from Chuck Wirtz and Randy Stark to All BWRVIP Committee Members, "Superseded "Needed" Guidance Regarding Crack Growth Assumptions," March 2012, SI File No. 1200283.205.
- 10. Tada, H., Paris, P., Irwin, G., The Stress Analysis of Cracks Handbook, 3rd. Ed., ASME Press.
2000. File No.: 1500270.301NP Page 12 of 18 Revision: 0 F0306-0IRI
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- 11. Hatch Nuclear Plant Unit 2 Core Shroud Drawings:
- a. Southern Nuclear Company Drawing No. 2-BN-6-6, Revision 3, "Core Shroud Weld Section and Details," SI File No. 1500270.203.
- b. Southern Nuclear Company Drawing No. 2-BN-6-4, Revision 1, "Shroud (Side View),"
SI File No. 1500270.203.
- c. Southern Nuclear Company Drawing No. 2-BN-6-5, Revision 6, "Core Shroud Weld Identification Roll Out (Inside View)," SI File No. 1500270.203.
- 12. Water Chemistry:
- a. Southern Nuclear Company Nuclear Management Guideline, NMP-CH-005-GLO I ,
Version 7.0, "E. I. Hatch Water Chemistry Strategic Plan," November 2014, Sl File No. 1500270.206.
- b. Southern Nuclear Company 41h Quarter 2014 Mitigation Health Report for Hatch Nuclear Plant Units 1 and 2, Sl File No. 1500270.206.
- 13. BWRVIP-62 Revision 1: BWR Vessel and Internals Project. Technical Basisfor Inspection Relief for BWR Internal Components with Hydrogen Injection. EPRI, Palo Alto, CA: 2011. 1022844.
EPRI PROPRIETARY INFORMATION.
- 14. Scarth, D., Wilkowski, G., Cipolla, R., Daftuar, S., and Kashima, K., "Flaw Evaluation Procedures and Acceptance Criteria for Nuclear Piping in ASME Code Section XI," Proceedings of 2003 ASME Pressure Vessels and Piping Conference, Cleveland, OH, Paper No. PVP2003-2026, pp. 45-61.
File No.: 1500270.301NP Page 13 of 18 Revision: 0 F0306-01 Rl
e Structural Integrity Associates, Inc.'* Table 1. HNP2 Core Shroud Evaluation Fluence and Failure Modes 50.1 EFPY Failure Analysis Method [4] Fluence, n/cm 2 [6] Mode [4] Non-Ductile LEFM with (( }} Fracture ff
),1 u
Table 2. HNP2 Shroud Bounding Axial Flaw Evaluation Results Service Level AlB Service Level CID Acceptable Axial Flaw at H4 Evaluation with bounding Acceptance Acceptance for 10 Years Method Results Results length at end of 10- Criterion Criterion (YIN) year interval = 15.5 K1=S.O K1<SO K1 = 12.6 K1<SO LEFM y in SF= 9.9 SF> 2.77 SF= 3.9 SF> 1.39 Note: K1 in ksi-in° 5
- File No.: 1500270.301NP Page 14 of 18 Revision: 0 F0306-01Rl
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;I ~1 -~ 1 Figure 2. Map of HNP2 core shroud ID indications based on 2R23 inspection results [1 ]. Annotations show approximate location, orientation, and length in inches.
File No.: 1500270.301NP Page 16 of 18 Revision: 0 F0306-0IRI
e Structural Integrity Associates, Inc.* tf pI* 2 F(~} * (I + 1.25~ )
'"' 0.6 o.t Figure 3. LEFM Solution for a Single Through-wall Axial Crack in an Internally Pressurized Cylinder [10]
File No.: 1500270.301NP Page 17 of 18 Revision: 0 F0306-0I Rl
e Structural Integrity Associates, Inc." Figure 4. LEFM Solution for an Infinite Array of Through-wall Cracks in a Plate with a Membrane Load [10] File No.: 1500270.301NP Page 18 of 18 Revision: 0 F0306-0IRI}}