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Lnservice Inspection Program Owner'S Activity Report for Outage 1R26
	ML14155A211
Person / Time
Site:	Hatch
Issue date:	06/03/2014
From:	Pierce C; Southern Co, Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-14-0840
	Download: ML14155A211 (52)
	v • d • e

{{#Wiki_filter:Charles R. Pierce Southern Nuclear Regulatory Affairs Director Operating Company. Inc. 40 Inverness Center Parkway Post Office Box 1295 Birmingham. Alabama 35201 Tel 205.992.7872 Fax 205.992.7601 SOUTHERN<<\ COMPANY JUN 0 3 2014 Docket Nos.: 50-321 NL-14-0840 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant- Unit 1 lnservice Inspection Program Owner's Activity Report for Outage 1R26 Ladies and Gentlemen: Enclosed is the ASME Section XI Code Case N-532-4 OAR-1 Owner's Activity Report for the 1R26 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 26 and are addressed in Table 2, "Abstract of Repair/Replacement Activities Required for Continued Service." This report is for the second period of the 4th Interval lSI activities (Interval 4, Period 3, Outage 1). This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369. Respectfully submitted, C. R. Pierce Regulatory Affairs Director CRP/RMJ/Iac

U. S. Nuclear Regulatory Commission NL-14-0840 Page2

Enclosures:

1. 1R26 Form OAR-1 Owner's Activity Report

2. 1R26 Form OAR-1 Owner's Activity Report, Table 1, Items with Flaws or Relevant Conditions that Required Evaluation for Continued Service

3. 1R26 Form OAR-1 Owner's Activity Report, Table 2, Abstract of Repair/Replacement Activities Required for Continued Service

4. Structural Integrity Associates Calculation 1200283.303 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bast, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President- Hatch Mr. B. L. Jvey, Vice President- Regulatory Affairs Mr. D. R. 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. R. E. Martin, NRR Senior Project Manager- Hatch Mr. E. D. Morris, Senior Resident Inspector- Hatch

Edwin I. Hatch Nuclear Plant - Unit 1 lnservice Inspection Program Owner's Activity Report for Outage 1R26 Enclosure 1 1R26 Form OAR-1 Owner's Activity Report

FORM OAR--I OWNER'S ACI'MTY REPORT Report Number 1-4-3-1 (Unit 1, 4111 Interval, 3RD Period, 1ST Report) Owner Southern Nuclear Operating Co, (as agent for Georgia Power Company), 40 Inverness Center Parkway, Binnins)!am, AL 35242 Plant Edwin I. Hatch Nuclear Plant, P. 0. Box2010, Baxley, Georsia 31513 Unit No. 1 Commercial service date t213Jn5 RefUeling outage no. (if applicable) ------ IR26 Cunent inspection interval Current inspection period Edition and Addenda of Section XI applicable to the inspection plans 2001 Edition with 2003 Addenda Date and revision of inspection plans HNP Unit 1 lSI Plan Volume 1 (Introduction), Version 3.0 7/1112011 - HNP Unit I lSI Plan Volume 2 (Components), Version 4.0 111512014 - HNP Unit I lSI Plan Volume 3 (Figures), Version 3.0 8/412011 - HNP Unit I lSI Plan Volume 4 (RPV), Version 6.0 2/412014 - HNP Unit I lSI Plan VolumeS (IWE), Version 6.0 21412014 - HNP Unit 1 lSI Plan Volume 6 (Pressure Testing), Version 7.0 21812014 Edition and Addenda of Section XI applicable to repair/replacement activities, if different than the inspection plans Same Code Cases used: N-460, N-513-3, N-532-4, N-663 (if applicable) CERTIFICATE OF CONFORMANCE I certify that (a) the statements made in this report are correct; (b) the examinations and tests meet the Inspection Plan as required by the ASME Code, Section XI; and (c) the repair/replacement activities and evaluations supporting the completion of IR26 confonn to the requirements of Section XI. Date CERTIFICATE OF JNSERVICE JNSPECfiON I, the undersigned, holding a valid commission issued by the National Board ofBoiler and Pressure Vessel Inspectors and tho State or Province of Georlria and employed by HSB Global Standards of Hartford. CT have inspected the items described in this Owner's Activity Report, and state that, to the best of my knowledge and belief. the Owner has perfonned all activities represented by this report in accordance with the requirements of Section XI. By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the repair/replacement activities and evaluation described in this report. Furthermore, neither the Inspector nor his employer shall be liable in any manner for any personal injury or property damage or a loss of any kind arising ftom or connected with this inspection.

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Edwin 1. Hatch Nuclear Plant - Unit 1 lnservice Inspection Program Owner's Activity Report for Outage 1R26 Enclosure 2 1R26 Form OAR-1 Owner's Activity Report, Table 1, Items with Flaws or Relevant Conditions that Required Evaluation for Continued Service

Poge J/J TABLEt ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Examination Category and Item Dt-scription E\*aluntion Description Item Number Stmcturnllntcgrity Associates calculation 1200283.303 shows that base During visual examinntion of the core shroud. four(4) material flaws, as well as flaws detected via augmented (BWRVIP) B-N-2/813.40 flaws were identified in base metal some distance !rom volumetric examination of core shroud welds arc acceptable stmcturnlly. anywdd. The subject SNC evaluation and SIA calculation arc attached within Enclosure 4. Engineering evaluation concluded that the support will be able to During visual examination of plant service wak-r withstand any additional bending stH:sses thut may be present from this support 11'41-SWH-801. it was discovered that the lug condition. The evaluation shows that the support has a safety factor of F-A I Fl.30 where the support rod is attached is bent slightly from 4.8. with a design load of 5090 pounds and an actual load of 1056 pounds. perpendicular. Titis evaluation concludes the support is not degraded, nor has the condition reduced the hangers design margin. Site design engineering performed a Non-Physical Design Change to During visual examination of residual heat removal detemtine more accurate hot and cold loads for the support. and the variable spring support IEII-RHRH-15, it was F-A I Fl.20C support detail drn\\ing was updated with the new hot and cold loads of discovered that the observed cold load of 5 158 lbs was 4455 lbs and 5055 lbs, respectively. The support was detennined to have outside of tht: 5% acceptuble rnng..:. been functional with the observed cold load. TI1is item was evaluated per code case N-513-3. and repaired during IR26

         ** I **

Plant service water 10" piping found below minimum with \VO SNC526583. This item is listed as the first line item on Table 2 wall thickness. of this report. ll1is item was evaluated per code case N-513-3, and repaired during I R26

         .. I **        Plant service water 30" piping through wall leak.       with WO SNC434160. This item is listed as th..: second line item on Table 2 of this rep(lrt.

This item was evaluated per code case N-513-3. and repaired during I R26

         .. I ..        Plant service water 6" piping through wall leak .       \\ith WO SNC516107. This is listed as the third line item on Tuble 2 of this report.

This item was evaluated per code case N-513-3. and repaired during IR26

         -I .*          Plant service water 6" piping through wall leak.        with WO SNC481420. This is listed as the fourth line item on Table 2 of this report.

Edwin I. Hatch Nuclear Plant - Unit 1 lnservice Inspection Program Owner's Activity Report for Outage 1R26 Enclosure3 1R26 Form OAR-1 Owner's Activity Report, Table 2, Abstract of Repair/Replacement Activities Required for Continued Service

Paplll TABLE2 ABSTRACf OF REPAIR I REPLACEMENT ACfiVJTIES REQUIRED FOR CONTINUED SERVICE Flaw or Relevaat Coadldoa Foaad Darlag Scbedalecl CodeCWs Item Descriptio* Dacrlpdoa or Work Date Complete Repair Replaeemeat Plaa Number Seccioa XI Eumlaadoa or Test? Plant Service Willa' - 10" Piping Performed weld inlay to restore 3 No 212112014 SNCS26S83 fOund below min wall thickness wall thickness Plant Savi= Water- 30" Piping 3 Repleced pipe sectiOII No 211712014 SNC434160 tJuu..wallleek I Plant Semce Walfl'- fl'/8" Pipins 3 Repleced pipe secti011 No 211312014 SNCSI6107 tJuu.walllcak Plant Service Water- (I' Piping tJuu.. 3 RepiDced pipe sectiOII No 2117/2014 SNC481420 wall leek

Edwin I. Hatch Nuclear Plant - Unit 1 lnservice Inspection Program Owner's Activity Report for Outage 1R26 Enclosure 4 Structural Integrity Associates Calculation 1200283.303

Southam Nuclear OoeratJng Company NMP-E6-018..QOS SOUIMUNA

          @MPANY Work Instruction                          Indication Notification         Verslon~J JMu-....-                                                                                   Page 1           I t-< ~uJ  ~-2CI~ 1'-/

Southern Nuclear Operating Company INF*Form*002wA Indication NotifiCation Fonn INF# I14H1007 P:ut 1 Flndlna!! Exam Date: NOE Math11d: lnstnldlaiii'RaviDev.: 02-11-2014 UT D PT 0 MT 0 ET o RT 0 vr X Other o NMP*E~24-206/11.1 Ccmmants: During IWI examination of the RPV Core Shroud Inner Oi;3meter areas, crack-like indications were found In the base material, away from the welds and heat affected zones (1811 1\H-123, H-127 & H-150).

Reference GEH INR H1R28 IWI-14-06. CR 775213 was written.

Date:: 02-21-2014 Dale: lltlo: s,r£" &JRVIP IS! Ei.ltr. Ccmmenb: Plo4DO 'oo attached NMP*ES-Ol0-200-FOl for ccmm~nts and evaluation. ~~~----------------- ~ Corract!ya Ac!lgn; No corrective Action required, please sea attached NMP-ES*010*200*F01 for further details. SlgnaiUIQ- Examiner: rJ J~ I SNT Level: N ~~ Part 4b- Addmlnng! Elcamjn!!tlons CScgpa Elg!!!m!aMl I£

     ~6 A'T'fr'CA-\.}) (\J-J\t.v..i\\)0*~ (17to7w'1).

ran :1o- SUTf?P"" Elc!m!na!!cm CFutym Exam Regu!mmen!sl Figure 1

Southern Nuclear Operating Company Nuclear NMP-ES-01 0-20Q-F01 Hatch Vessel & Internals Program - Flaw SOUtHERN.\_ Management Version 1.0 COMPANY Evaluation Form c.......-~..., Form Page 1 of 7 BWRVIP EVALUATION TE# 776767 Component Core Shroud Applicable NMG NMP-ES-01 O-GL02 BWRVIP Guideline# _ _....:7~6~- INF/INR # H1R261WI-14-06/114H1007 Condition: During IWI activities associated with the Edwin I. Hatch Nuclear Plant refueling outage 1R26 in February 2014. inspection personnel performing scheduled visual VT-3 examinations of the Unit 1 Core Shroud reported new crack-like indications not associated with a horizontal or vertical weld. The indications were reported to SNC via INR-IWI 06 and are located at 80°, 105° and 350° azimuths and between horizontal welds H3 and H4 on the inner diameter surfaces of the shroud. A total of thirteen 1O*degree lanes representing >36% of the shroud surfaces between H1 and H5 were examined. Indications were reported at the 80°, 105° and 350° azimuths between the elevation of horizontal welds H3 and H4 but not associated with any shroud horizontal or vertical weld or weld heat affected zone CHAZ). The indication detected at 80° was detected from cell location 5Q-35. Additional characterization of this indication was performed employing VT-1 resolytjon and volumetric (UT) examination. The indication was measured to be 2.9n in length with a maximum deeth of 0.52n. It is characterized as primarilY in the vertical direction with a short (=0.25") horizontal component. Strong visual evidence of surface grinding was also determined to be present. Two indications were also detected at azimuth 350°. One indicatjon was measured to be about 2.8" long and the other approximately 5116" and the two are in close proximity to each other. Surface gririding was also present at these indications. The final indication reported was at azimuth 105°. This indicatjon was reported to be approximately 3" in length. Surface grinding is also aoparent from VT-3 video captures from this Inspection. This INA covers flaws located within the base material. away from any welds. Raws associated with Shroud OD VT-3 inspections are documented on INF 114H1011. Raws associated with the vjsual inspection of the V4/H4 weld intersection are documented on JNF 114H1012. Inspection results of the Shroud 10 VT-3 can be found on INF 114H1014. Results from the visual examination of the V8/H5 weld intersection are documented with INF 114H1015. INF 114H1016 documents the evaluation of all shroud UT examination.

Southern Nuclear Operating Company Nuclear NMP-ES-0 10-200-F01 SOUTHERN COMPANY A Management Hatch Vessel & Internals Program- Flaw Evaluation Form Version 1.0 E""V :.Stnw t;.*~JWU' Form Page 2 of 7 Disposition: These indications are most likely a result of surface cold work (grinding). As depicted in the referenced GE INR document. grinding marks can be seen in the vicinity of all indications. UT data was collected on one of these indications and documented within CNF-SHRD-003, indicating a length of 2.9". and a depth of 0.52". Structural Integrity Associates calculation 1200283.303 evaluates these indications and determines that they are bounded structurally by larger flaws elsewhere in the shroud. UT data taken on one of the subject flaws. shows that jt is not through-wall. The remaining flaws are very similar and are likely similar in depth. However. if the flaws are through wall. they would still be bounded by the 1200283.303 evaluation and would be acceptable for continued operation. Basis for Disposition: These new indications were evaluated within SIA report 1200283.303. As stated within that document. because of the newly identified flaws' material properties. loads and assumed crack growth rates are similar to those elsewhere in the shroud, we can compare them with other flaws in the shroud. The flaws within the base metal are not structurally significant compared with the limiting flaw in the 1200283.303 calculatjon. All indications evaluated within SIA report 1200283.303 were structurally assumed to be through-wall and therefore the base metal indications are bounded by the evaluation regardless of depth and are acceptable for continued operation. This evaluation was done in accordance with BWRVIP-76-A. which is consistent with ASME Section XI rules with the exception that material properties are more conservatively accounted for with respect to irradiation effects. Scope Expansion (ref. NMI step 4.6.9) See attachment discussing scope expansion. Follow-on or Supplemental Examinations (ref. NMI step 4.6.10) See attachment discussing supplemental examinations. References (ref. NMI step 4. 7.1) H1 R26 IWI-14-06 Core Shroud Plate Indications. Indication Notification Report. Februarv 2014

Southern Nuclear Operating Company Nuclear NMP-ES*01 0*200-F01 SOUTHERN COMPANY A Management Hatch Vessel & Internals Program- Flaw Evaluation Form Version 1.0 lrf~JDr.Suwti.r.,... Form Page 3 of7 SIA 1200283.302 Hatch Nuclear Power Plant Unit 1 Core Shroud Vertical Welds V5N6 Crack Growth and Fracture Mechanics Evaluation. February 2014 SIA 1200283.303 Hatch Nuclear Power Plant Unit 1 Core Shroud Axially Oriented Flaw Evaluation and Leakage Rate Calculation- H4 Weld. Base Material Flaws between H3 and H4 Weld. V7 Flaws. VS Flaws. Februarv 2014 CNF-SHRD-003 Base Metal Flaw between H3 and H4 at -65° azimuth. February 2014 BWAVIP-76-A BWR Core Shroud Inspection and flaw Evaluation Guidelines. November 2009 NMP-ES-010-GL02 Version 9.0 BWRVIP Core Shroud and Shroud Stabilizer Guideline. April 2013 ASME Section XI 2001 Edition with 2003 addenda Review and Approval (ref. NMI Step 4.7) Responsible Engineer: Andrew Gordon/~~ Date: o 1 Aorll. 2014 Supervisor Approval: "D41* MtswrJ/m4v.:u,.J Date: '1-~4'1

Southern Nuclear Operating Company Nuclear NMP-ES*010*200-F01 SOUTHERN COMPANY A Management Hatch Vessel & Internals Program- Flaw Evaluation Form Version 1.0 £uwuhl~y....~* Form Page 4of 7 Edwin I. Hatch Nuclear Plant- Unit 1 Technical Justification Regarding Scope Expansion of ASME Code Section XI 8-N-2 Components Conclusion It can be concluded that additional examinations are not warranted because any potential additional accessible areas have already been adequately inspected via various combinations of VT-3, VT-1, and UT.

Background

During IWI activities associated with the Edwin I. Hatch Nuclear Plant refueling outage 1R26 in February 2014, inspection personnel performing scheduled visual VT-3 examinations of the Unit 1 Core Shroud reported new crack-like indications not associated with a horizontal or vertical weld. The indications were reported to SNC via INR-IWI 06 and are located at aoo, 105° and 350° azimuths and between horizontal welds H3 and H4 on the inner diameter surfaces of the shroud. A total of thirteen 10-degree lanes representing >36% of the shroud surfaces between H1 and H5 were examined. Indications were reported at the 80°, 105° and 350° azimuths between the elevation of horizontal welds H3 and H4 but not associated with any shroud horizontal or vertical weld or weld heat affected zone (HAZ). The indication detected at 80° was detected from cell location 50-35. Additional characterization of this indication was performed employing VT-1 resolution and volumetric (Un examination. The indication was measured to be 2.9" In length with a maximum depth of 0.52". It Is characterized as primarily in the vertical direction with a short (=0.25") horizontal component. Strong visual evidence of surface grinding was also determined to be present. Two indications were also detected at azimuth 350°. One indication was measured to be about 2.8" long and the other approximately 5/16" and the two are in close proximity to each other. Surface grinding was also present at these indications. The final indication reported was at azimuth 105°. This indication was reported to be approximately 3"1n length. Surface grinding Is also apparent from VT*3 video captures from this inspection. These examinations were performed to comply with ASME Code Section XI IWB 25Q0-1, Category 8-N-2 Item Number B13.40 "Core Support Structures" requirements which require that the surfaces of accessible core support structures be examined using visual VT-3 techniques. Because planned fuel cell evacuations made portions of the shroud inner diameter accessible these examinations were scheduled. The rules of IWB-2420 usuccessive Inspections" and IWB-2430 Additional Examinations" apply because these are ASME Code Section XI examinations. This Technical Justification Regarding Scope Expansion for Examination has been prepared to determine the requirements and logic for compliance with these "Additional" and "Successive" requirements.

Southern Nuclear Operating Company Nuclear NMP-ES-01Q-200-F01 SOUtHERN A C....,R _ _ _ COMPANY Management Hatch Vessel & Internals Program - Flaw Evaluation Form Version 1.0 Form Page 5 of 7 Additional Examinations IWB-2430(a) states that "examinations performed in accordance with IWB-2500-1, **. , that reveal flaws or relevant conditions exceeding the acceptance standards of Table IWB-341 0-1 shall be extended to include additional examinations during the current outage. The additional examinations shall include an additional number of welds, areas, or parts, included in the inspection item, equal to the number of welds, areas, or parts, included in the inspection item that were scheduled to be performed during the present inspection period. The additional examinations shall be selected from welds, areas, or parts of similar material and service." The original Code Section XI requirement for B-N-21tem 13.40 is "Accessible Surfaces" employing visual VT-3 examination techniques. In order to assess potential locations 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 that way. The shroud surfaces which are considered "always" accessible are those surfaces which can be examined from the outer diameter of the shroud because the only obstructions are permanent ones such as jet pumps or tie rods. For record-keeping these items are scheduled throughout the interval by lanes corresponding to 10 degree wide azimuths. Unlike the outer diameter, the Inner diameter of the shroud along with any surfaces below the core plate do not normally 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 current outage scope contains those azimuthal lanes that were to be evacuated and that were NOT previously examined in the 4th 10-year lSI Interval. There are three lanes that fall into this bin, I.e. that were previously examined during the interval AND adjacent to evacuated fuel cells. It is not an ASME Code Section XI requirement to evacuate fuel cells or remove other components to "make" surfaces accessible. Satisfying the requirement that;

The additional examinations shall be selected from welds, areas, or parts of similar material and service.** n requires some judgment as to what constitutes similar material and service. The judgment in this assessment is that detected cracking is due to original surface cold work i.e. grinding combined with irradiated material from high fluence levels. Thus similar service would include the locations with similar fluence, combined with the potential for surface cold work.

The three additional azimuthal lanes from the inside diameter have surface fluences corresponding to locations which experienced the indications and therefore should be considered for "additional examination~>. Since the degradation effect is cracking, any examinations for cracking with sufficient coverage in those lanes already conducted or scheduled would be considered acceptable for meeting the intent of the Additional Examination requirements. From an elevated fluence perspective it can be shown that the areas where the flaws were detected are in the more highly irradiated areas of the shroud inside diameter. Conservative similar areas would be those corresponding to the "beltline" of the shroud inside diameter between horizontal weld H3 and horizontal weld H5. The lanes to be considered for additional examinations are: Azimuth 45°-55° lane {cell) 46-43 Azimuth 225°*235° lane (cell) 06-11

Southern Nuclear O~eratlng Com~any Nuclear NMP-ES-Q1 0-20Q-F01 SOUTHERN COMPANY A Management Hatch Vessel & Internals Program - Aaw Evaluation Form Version 1.0 h'IJI*Sww thrlaril~ Form Page 6of7 lane (cell) 10*47 These cells were evacuated and examinations performed during 1R24 in 2010. No indications in the base material were recorded although the expected indications near weld heat affected zones were recorded and appropriately evaluated under BWRVIP guidance. Lane 46-431s adjacent to vertical welds VS and V9, lane 06-11 is adjacent to vertical weld V6 and lane 10*47 is adjacent to vertical welds V4 and va. Two of the indications that were detected were in areas outside the scheduled "lanesn. The 350 degree indication was 5 degrees outside the width of the 335"-345" lane. The 80 degree indication was 5 degrees outside the 65°-75° degree lane. Since the lanes adjacent to each side of all three lanes in consideration for additional visual VT-3 examination were performed, it is logical to state that there is sufficient field of view overlap to conclude these surfaces have been adequately examined via visual VT-3. As additional conservatism for the adequacy of examinations of the shroud can be demonstrated by noting those examinations performed during 1R26 by method other than visual VT-3: Examinations in lane 46-43: V5 UT from 10 (UT this weld is in lane 46-43) V9 UT from 00 (weld Is below core plate thus Inaccessible from 10) Exams in lane 06-11: V6 UT from 10 (this weld is in lane 06-11) Exams in lane 10*47: V4 EVT-1 from 10 and UT from 10, this weld is located in lane 10-47 va EVT-1 and UT from 10 (this weld is located in lane 10*47) Other relevant shroud exams in the beltline: V7 UT from 00 (inaccessible 10 lane and lower than elevation of detected flaws) V3 UT from 00 (inaccessible 10 lane and at elevation of detected flaws) V10 UT from 00 (weld is below core plate and thus inaccessible from 10) In summary, the 3 areas considered for additionaJ examinations have already been assessed via 1R26 scheduled exams in the same lanes, which include visual and voiumetric exams. In addition, portions of two lanes adjacent to lanes with indications noted were also examined via VT-3 during 1R26. Based on this assessment, additional visual VT-3 examination of the three accessible lanes available during 1R26 is not warranted to meet Code requirements. Successive Inspections The applicable requirement for compliance regarding successive examinations is IWB-2420(b) whi.ch requires that "the areas containing the flaws or relevant conditions shall be reexamined during the next three Inspection periods listed in the schedule of the Inspection program of IWB-2234." IWB-2420(c) prescribes that "If... the flaws or relevant conditions remain unchanged the component examination may revert to the original schedule of successive Inspections." This outage is the first outage of Period 3 of lSI interval4. Compliance thus dictates that follow-on examinations be conducted during an outage in each of the three periods of the 51h Interval. The B-N-2 caveat allowing only accessible surfaces does not

Southern Nuclear Operating Company Nuclear NMP-ES-01 0-200-F01 Hatch Vessel & Internals Program- Aaw SOUTHERN..\ Management Version 1.0 COMPANY Evaluation Form ~,.,.....,...,.,......,.. Form Page 7 of 7 apply meaning that evacuation of the applicable cells to allow access must be scheduled to meet ASME successive examination requirements. Extent of Condition For extent of condition it is recommended that In a future outage, possibly in conjunction with the successive examinations, Plant Hatch should schedule some additional surface examinations In base material, particularly in areas that have not been examined during the 10-year lSI interval. Additional examinations will be scheduled through the Hatch Unit 1 lSI Plan documents, and is being further tracked byTE 780835.

lJ Strocturallntegrity Associates, Inc." File No.: 1200283.303 Project No.: 1200283 CALCULATION PACKAGE Quality Program: (gl Nuclear 0 Commercial PROJECT NAME: Hatch V5N6 Shroud Vertical Weld Evaluation CONTRACT NO.: 28566, Rev. 0, Purchase Order: SNG 10062021, Rev. 0 CLIENT: PLANT: Southern Nuclear Operating Company, Inc. Edwin I. Hatch Nuclear Plant, Unit l CALCULATION TITLE: Hatch Nuclear Power Plant Unit I Core Shroud Axially Oriented Flaw Evaluation and Leakage Rate Calculation- H4 Weld, Base Material Flaws between H3 and H4 Weld, V7 Flaws, V8 Flaws Project Manager Prcparcr(s) & Document Affected Re,*ision Dcscri1)tion Approval Chcckcr(s) Pages Si1matnrc & Date ~:. ."'" & Date 0 1 - 21 Initial Issue Responsible Engineer A A-9 Daniel V. Sommerville 27FEB2014 Daniel V. Sommerville 27FEB2014 Responsible Verifier Raju Ananth 27FEB2014 14, 15, 16 Added leakage rate a _L} /J1 Responsible Engineer

                                                            ~*---YF                    ~

A-10-A-14 results for 2 cycles of operation Daniel V. Sommerville 3APR2014 Daniel V. Sommerville 3APR2014 Raju Ananth 3APR2014 Page 1 of21 F0306-01Rl

Table of Contents

1.0 INTRODUCTION

......................................................................................................... 4 2.0     OBJECTIVES ................................................................................................................ 4 3.0     METHODOLOGY ........................................................................................................ 4 3.1       Flaw Evaluation ................................................................................................. 4 3.1.1 Characterize Fla~vs ............................................................................................ 5 3.1.2 Material Properties ........................................................................................... 5 3.1.3 Inspection Uncertainty ....................................................................................... 5 3.1.4 (~rack (iro~vth     .................................................................................................... 5 3.1.5 Fracture Mechanics ........................................................................................... 6 3.2       Leakage Calculation .......................................................................................... 7 3.2.1 Crack Opening Area .......................................................................................... 7 3.2.2 Determine Crack Distriblllion ........................................................................... 7 3.2.3 Leakage Rate and Loss Coe.fficienl Calculation ................................................ 8 4.0     DESIGN INPUTS ........................................................................................................ tO 5.0     ASSUMPTIONS .......................................................................................................... l2 6.0     CALCULATIONS ....................................................................................................... l4 7.0     RESULTS .................................................................................................................... l5

8.0 CONCLUSION

S ......................................................................................................... 15

9.0 REFERENCES

............................................................................................................ 16 APPENDIX A CALCULATIONS ....................................................................................... A-I File No.: 1200283.303                                                                                                                    Page 2 of21 Revision: 1 F03()6.() I R I

t)Sttuc:llii'BIIIIflgrlly Associates, Inc.* List of Figures Figure I. HNP I Core Shroud Configuration [ 12] .................................................................. 18 Figure 2. Previous HNP I Core Shroud Inspection Results [ 12] ............................................ 19 Figure 3. LEFM Solution for a Single Through-wall Axial Crack in an Internally Pressurized Cylinder ............................................................................. 20 Figure 4. LEFM Solution for an Infinite Array of Through-wall Cracks in a Plate with a Membrane Load ................................................................................. 21 File No.: 1200283.303 Page 3 of21 Revision: 1 F031J6.0JRI

l)stnmtumtiRfegrlfy Associates, lfic.*

1.0 INTRODUCTION

Multiple reportable indications were detected in the Hatch Nuclear Plant, Unit I (HNPI) core shroud during the Spring 2014 refueling outage planned inspections [1]. Numerous axially oriented indications were reported on the outside and inside surfaces of the core shroud in:

The vicinity of the core shroud circumferential welds H4 and H5
The vicinity of the vertical weld V4 intersection with circumferential weld H4,
The core shroud base material, between the core shroud H3 and H4 circumferential welds,
The heat affected zones (HAZ) of the V5, V6, V7, and V8 welds.

Figures I and 2 illustmte the HNP I core shroud configuration, weld locations, and identify historically reported indications. 2.0 OBJECTIVES The objectives of the work documented in this calculation package are to: I. Perform a flaw evaluation and leakage calculation for all ofthe axially oriented indications, on the inside and outside surfaces of the core shroud, reported during the Spring 2014 HNPI refueling outage, with the exception of the structural evaluation of the VS and V6 welds which is addressed in Reference [19],

2. Perform a combined leakage calculation for all reported through-wall indications and assuming a conservative condition around the uninspected region of the H4 weld.

3. Perform a flaw evaluation of the base material indications in accordance with the requirements of the American Society ofMechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code, Section XI rules for in-service inspection of light water reactors. Circumferentially oriented indications are not evaluated in this calculation package since the tie rod shroud repair installed at HNP I structurally replaces the circumferential welds [ 12]. 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, Rev. I [2]. The following process is used: I. Characterize location and dimensions of the flaws,

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.: 1200283.303 Page 4 of2l Revision: l 1'0306-0IRI

eSIIvc:lumiiRIJgrlty AssociatrJs, Inc.*

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 [2] are based on the rules of ASME Code, Section XI IWB-3600 and non-mandatory 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 methodolob'Y have been reviewed and accepted by the U.S. Nuclear Rebrulatory Commission (NRC) with respect to inspection and evaluation of the Boiling Water Reactor (BWR) core shroud [2]. Each step of the process is described separately below.

3. I. I Characteri=e 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 (GEH)

[I]. 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. All flaws are treated as through-wall for the flaw evaluation. A single bounding flaw is selected for evaluation rather than individually evaluating every flaw reported in all locations on the core shroud. 3.1.2 Material Properties BWRVIP-100-A [3] is used to identify the failure modes appropriate for the level offluence 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-A [3] and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section II, Part D [4a] are also used to select the appropriate tensile properties for the fluence level and temperature. BWRVIP-100-A [3] is used to select the appropriate fracture toughness for the shroud fluence. The bounding fluence along the entire core shroud height at 49.3 effective full power years (EFPY) [5] is conservatively used for this evaluation. The current refueling outage is lR26 and the end of design life EFPY is reported to be 49.3 EFPY in Reference [5], which corresponds to refueling outage 1R36.

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 [6].

3. I .4 Crack Growth Consistent with the methods provided in BWRVIP-76, Rev. I [2] and the clarifying guidance given in Reference [7] intergranular stress corrosion crack growth (IGSCC) is calculated for the evaluation 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.

File No.: 1200283.303 Page 5 of21 Revision: 1 F0306-0IRI

eSinn:tumlllllsgrlly Associates, Inc.* The IGSCC length crack growth rate provided in BWRVIP-76, Rev. 1 [2] and BWRVIP-14-A (8] is used for all flaws: da =5.0 . 10 _5

in dt hr

3. 1.5 F'racfllre Mechanics Stability of the core shroud is assessed by performing both limit load and linear elastic fracture mechanics (LEFM) calculations. The limit load calculations evaluate stability of the core shroud structure; whereas, the LEFM calculations evaluate stability of the crack.

The ASME B&PY Code, Section XI, non-mandatory Appendix C limit load method [4b] for through-wall axial flaws is used since the method provided in BWRVIP-76, Rev. 1 [2] 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, non-mandatory appendix C solution is considered a more appropriate solution. The following equation is specified in Section C-5410 for through-wall axial flaws [4b]:

                             !~)---;., .

2 lallo,.. = l.58*v"m (- U flmr )

                                              -      -I                                                (1)

Uuaap Where fallow 2a, allowable flaw length, in. Rm = Shroud mean radius, in. LEFM solutions published in Reference [9] are selected for this evaluation. The solutions used for this evaluation are consistent with those suggested in BWRYIP-76, Rev. 1 [2]. The following LEFM solutions are used:

Single through-wall axial flaw in an internally pressurized cylinder (See Figure 3) [9, pg. 485],
Infinite array of parallel through-wall flaws in a plate subjected to a membrane load (See Figure 4) [9, pg. 256],
Plastic zone size correction as given below [9, 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. File No.: 1200283.303 Page 6 of21 Revision: 1 F0306..()JRI

  !S)Sttucturalllllegrlty AssaclatBs, Ina.*

The radius of the plastic zone size is added as an additional crack len~:,rth at each end of each flaw. The plastic zone size correction is estimated, for conditions of small scale yielding, using the following equation [9, pg. 16]: (2) Where a. is used to adjust for plane strain or plane stress conditions at the crack tip, where: Plane Strain: a.=l/61t Plane Stress: a.= ll21t CJvield 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. For leakage rate calculations a flaw specific plastic zone size is calculated to remove unnecessary conservatism. 3.2 Leakage Calculation The leakage calculation is performed using methods consistent with those given in BWRVIP-76, Rev. I [2]. The leakage calculation is performed using the following process:

l. Calculate end of interval crack opening area (COA),

2. Determine crack distribution,

3. Calculate through wall leakage.

Each item is addressed separately below. 3.2.1 Crack Opening Area The COA is conservatively determined assuming a rectangular slot of total opening area as given in the LEFM solution presented in Figure 3. 3.2.2 Determine Crack Distribution Current industry methods applicable to the core shroud provided in BWRVIP-76, Rev. I [2] do not require that through-wall flaws be postulated to exist in uninspected regions of the core shroud. To be conservative, this calculation includes an assumption that a distribution ofthrough-wallleaking flaws exists around the entire circumference of the core shroud of the same density (number and size) as reported in the inspected region for the flaws reported in the V4/H4 intersection. File No.: 1200283.303 Page 7 of21 Revision: I F0306-01RI

  !l)Sintc:lumiiRtegrlly Assoclat9s. Inc.*

3.2.3 Leakage Rate and Loss Coefficient Calculation The volumetric leakage rate through a through-wall crack can be determined using standard 1-D fluid dynamics methods [10]. The Bernoulli equation, applicable to steady, incompressible, frictionless flow along a streamline, is modified by incorporating a minor loss term consistent with basic fluid mechanics. Next, published minor loss coefficients for flow through an abrupt contraction, to simulate flow entering the crack, and an abrupt expansion, to simulate flow leaving the crack, are selected. Equation (4) shows the Bernoulli equation with the minor head loss term included: (4) Where: g is the acceleration of gravity. inlsec2 gc is the proportionality constant= 386 lbm-in/lbf-sec2 HL is the minor loss, in 2/sec 2 PI is the pressure inside the core shroud, psi p2 is the pressure outside of the core shroud, psi VI is the fluid velocity normal to the crack opening area inside the shroud, in/sec v2 is the fluid velocity normal to the crack opening area as it leaves the core shroud and enters the annulus, in/sec Z1 is the elevation at the entrance to the crack, in Z2 is the elevation at the exit of the crack, in p is the mass density of the leaking fluid, lbrnlin3 Taking the elevation of the crack entrance and exit to be equal enables the elevation terms to cancel each other out in Eq. (4). Further, recognizing that the fluid inside the core shroud is either stagnant or moving parallel to the plane defining the crack opening, the V1 term is also zero. This allows Eq. (4) to be simplified to:

              ~ -P,     v.,2 H gc    p - =    i    + L Next, the minor loss term Ht is expressed as:

v.,2 HL = 2(Kcom. + K£.11, ). Where: Kconl. is the loss coefficient for sudden contraction is the loss coefficient for sudden expansion File No.: 1200283.303 Page 8 of2l Revision: I F0306-0IRI

Recognizing that the term P 1-P2 is the pressure difference across the core shroud, M, and inserting the equation for the minor loss allows Eq. (4) to be simplified to the following expression for the velocity of the leaking fluid as it exits the crack: V, = 2gc .t:J>

            -      P * {1 + Kcont + KExp.)

Finally, introducing the definition ofvolumetric flow as Q=AV, where A is the COA, and applying an additional algebraic manipulation to separate the loss term gives the following expression for the volumetric leakage rate through the crack: Q-

              - ~{1 + Kcont I
                                + KE.tp).

A

lg,*M' P (5) Thus, the loss coefficient defined in Eq. (5) above is given by: Recognizing that the ratio of the COA to the internal flow area of the shroud will approach zero for most crack sizes of practical importance, and reCO!,'llizing that the leakage flow exiting the crack can be acceptably treated as entering an infinite reservoir, the following loss coefficients are obtained from Reference [10 pg. 484]: Sudden contraction: 0.5 Sudden expansion: 1 Thus, 1 C= =0.63

                .J{2.5)

A loss coefficient of0.63 is applied in the leakage rate calculations for this calculation package. Finally, the equation to be used to calculate the volumetric leakage flow rate through each crack configuration considered in this evaluation is given by Equation (6) below: Q =0.63*A* ~ 2 g'";t),.P (6) This equation is consistent with the equation given in BWRV1P-76-A [2]. Total leakage rate can be calculated as the sum of the leakage rates calculated for each crack. File No.: 1200283.303 Page 9 of21 Revision: 1 F0306-0IRI

  ~Stnrclllrall1118grlly Associatfls, Inc.*

4.0 DESIGN INPUTS The design inputs used for this calculation are identified below: Geometrv: The shroud geometry is taken from Reference [II ] :

Shroud 10: 174.5 in. [II a]
Shroud Thickness, t: 1.5 in. [I I b]

Loads and Through-wall Stress Distributions: The upper shroud RIPD values are taken from Reference [ 12], and are summarized as follows:

Level A RIPD: 7.73 psi
Level B RIPD: I 1.59 psi
Level C RIPD: 29.5 psi
Level D RIPD: 29.5 psi IGSCC Crack Growth Rate:
The length CGR: 5 x Io-s inlhr [7, 8].

Reactor Coolant Water Chemistrv: HNPI implemented HWC in September 1987, and started Noble Metal Chemical Addition (NMCA) in March of 1999 [12, 13]. Under HWC conditions and NMCA, the shroud horizontal welds H3, H4 and H5 are considered mitigated [14, Table 4-1]. Shroud Fluence: The peak core shroud fluence at 49.3 EFPY is: 4.01 x to-"I nlcm-' [5] Material Type: The shroud material is SA-240 TP304 stainless steel [lla]. Material Properties: For fluence values greater than 3 x 1021 nlcm2, LEFM analysis should be used to determine the structural stability of flaws in the core shroud, and a static initiation, plane strain, mode I fracture toughness: 50 ksivin [3]. File No.: 1200283.303 Page 10 of21 Revision: I F0306-0IRI

l')SituctumlllfiJgrlty Assoc/atts, Inc.* The material flow stress and yield stress both increase with fluence [3]. 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 [4a]:

cru (un-irradiated, 550 oF): 63.4 ksi
cry (un-irradiated, 550 OF): 18.9 ksi
crr(un-irradiated, 550 *F): 41.2 ksi (taken as the average ofcru and cry)
Elastic Modulus of Type 304 Stainless Steel at 550° = 25.55 x 106 psi
Poisson's Ration of Type 304 Stainless Steel at 550° = 0.3 For the leakage rate calculations the fluid density is:
Coolant mass density, p: 62.4 lbrn!ft', water at 50 °F [15]

Initial Flaw Distribution: The flaw lengths, depths and distributions are taken from the 2014 INRs [I]. Since the reported number of indications is large and many have been detected using a VT-3 inspection technique (accuracy of sizing is uncertain), no attempt is made to tabulate 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 ultrasonic testing (UT) inspection data are taken from the applicable demonstrations for the inspection tooling identified in the INRs. The evaluation factors are taken from BWRVIP-03 [6a] and BWRVIP Letter 2014-015 [6b]. Bounding length evaluation factors are taken from the applicable demonstmtions. No depth evaluation factors are used since all flaws are treated as through-wall. Since most visual examinations were VT-3 there is no applicable length evaluation factor. This lack ofVT-3 sizing uncertainty is addressed by clearly showing that the flaw size selected for evaluation bounds all visually reported indications. The following length evaluation factors are used:

TEIDE Tool UT Length Evaluation Factor: 0.149 inches [6b, Demonstration 73]
Shroud OD UT tool Length Evaluation Factor: 0.137 inches [6a, Demonstration 47]

Operating Cycle Duration: HNP l is on a 2 year operating cycle [ 12]. File No.: 1200283.303 Page II of21 Revision: I F0306-0IRI

l)Stnlclumllllt8grlly Assoclatls, 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 a.\:mmption is appropriate for the flaws reported to he through-wall and conservative for the flaws reported to be less than through-wall. This assumption provides a bounding flaw evaluation for all flaws.

2. The 49.3 EFPY peak fluence is used to determine the fracture toughness for all flaws.

This assumption is conservative because it applies the maximum fluence projected at the end of the operating license to all locations. The cun*ent refiteling outage is 1R26 and the end ofdesign life EFPY is reported to be 49.3 EFPY in Reference [5/, which correspond<; to refiteling olllage JR36.

3. A 100% capacity factor is assumed for crack growth.

This assumption is conservative because it uses the largest number of hours possible, each year, for crack growth.

4. Flaws are allowed to grow through the horizontal welds.

Inspection data from the HNP 1 core shroud shows evidence offlaws growing through the weld [ 1}; therefore, this assumption is considered appropriate.

5. Uninspected regions of welds are assumed to be 100% cracked through-wall.

This assumption is appropriate because it conservative{v assumes that the entire length of each uninspected region is flawed. BWRVIP-76, Rev. 1 [2} provides guidance for using a statistical approach to determining the percent of uninspected regions to be assumed flawed. This approach could be used here. HoWL'Ver, since the uninspected regions are relatively short, the crack growth in the length direction would extend the length of these regions before the EO!.

6. A single bounding flaw is evaluated in this calculation package which is defined to bound the length reported for all axial indications.

Rather than £.'Valuate all axial(v oriented indications separately, a single flaw evaluation is peiformed of the single largestflaw. It is shown in this calculation that parallel flaws are bounded by a single flaw; thus, this approach hounds all reported flaw lengths and configurations (single or multiple paral/el.flaws)

7. The longest flaw at the H4/V4 intersection is assumed to have a length which is composed of the total flaw length reported in the inspection window (8.0 inches) plus an additional4.0 inches to account for the fact that the flaw extended to the top side of the inspection window. In other words, the upper end of the flaw was not captured by the inspection.

This assumption is considered to he conservative since the flaw length is increased by an additional 50% to accoumfor existence ofadditional cracking above the inspection window. Further, the inside smface EVT-1 examinations peiformed at the H4/V4 intersection [1c} show File No.: 1200283.303 Page 12 of21 Revision: I F0306-0JRI

  !S}Situctutallnlflgrtty Associatfls, Inc.*

that all axially oriented indications showed JD swface lengths of less than 5.5 inches which supports the assertion that using an assumed length of 12.0 inches is conservative.

8. Leakage rates are calculated using a lower bound fluid temperanue of 50 °F.

This assumption is considered conservative since residual heat in the core and core shroud will heat the emergency core cooling flow that enters the core which will raise the fluid temperature. This value is considered to be a conservative lou:er bound on the temperatllre of the fluid that can he routed to the core during ECCS operation. The colder the fluid, the larger the fluid density will be, which will result in larger leakage flow rates.

9. The crack opening area shape is rectangular.

The literature supports an elliptical COA [16, 17/. Consequently, assuming a rectangular COA is conservative.

10. Ratio of the COA for each crack to the flow area inside the shroud is zero.

1he crack opening areas for each crack are on the order ofa fraction ofa square inch; whereas, the flow area of the 175 inch mean diameter shroud is orders ofmagnitude larger. Consequently, this assumption is appropriate. II. The annulus between the RPV and core shroud is effectively an infinite reservoir.. The same rationale as applied to Assumption 10 above is applicable to this assumption; therefore, this assumption is reasonable.

12. Flow is steady, incompressible, frictionless and single phase.

This assumption is reasonable since:

a. The leakage flow is expected to reach quasi-steady state on a time scale on the order of 1 second or less; hence the flow is steady,

b. lhe leakage velocities are predicted to be much less than the local sound speed of water at 50 "F; hence the incompressible assumption is just{fled,

c. Ignoring viscosity effects inthe.flow is conservative since consideration offriction e.ffects would tend to reduce flow velocities and result in smaller leakage rates.

d. 11w leakage rates calculated are for long term post LOCA when the vessel has depressuri:ed; therefore, the coo/am remains single phase.

13. The difference in the elevation of the crack entrance and exit is negligible.

111is is a reasonable assumption since the shroud is 1.5 inches thick and the change in elevation across the shroudfor a crack whose growth is driven by stresses which would tend to keep the crack in an orientation norma/to the shroud swjace is negligible. File No.: 1200283.303 Page I3 of21 Revision: I FOJ06-0JRI

l}Sinlc:tumlllltegrlty Associates, Inc.*

14. Since the entire circumference of the shroud weld H4 was not inspected and since four through wall flaws are reported in the one inspection window at this elevation the flaw density reported in this window (number and size) is assumed to be present around the entire circumference of the shroud at this elevation.

This assumption is considered to be more conservative than curre1111y required in indwmy guidance pertaining to the BWR core shroud /2/ yet also reasonable considering the lack of volumetric inspection coverage at this elevation, number of through-wall reported indications, and extent ofcracking elsewhere on the HNP I core shroud. 6.0 CALCULATIONS All calculations are performed using the MathCAD software [ 18]. The calculations performed for this flaw evaluation are contained in Appendix A. Flaw stability is assessed by performing bounding calculations for all reported indications, as follows: I. The bounding shroud fluence at 49.3 EFPY is applied for all shroud elevations resulting in a lower bound fracture toughness for all flawed locations.

2. Review of the LEFM solution shown in Figure 3 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 F11 curve for s-o which corresponds to a single flaw. Consequently, multiple aligned flaws can be bounded by treated each as a separate flaw.

3. The longest flaw reported in the H4N4 intersection is selected for evaluation and clearly bounds all flaw lengths reported in the V7, V8, and base material. This flaw is assumed to be a single continuous flaw passing through the uninspected regions adjacent to the H4 weld and through the weld itself. The flaw is further increased by -50% to account for the fact that the UT examination did not capture the crack tip on the upper side of the inspection window. The initial flaw size considered, before additional ofuncertainty and crack growth, is 12 inches.

Leakage rates are calculated assuming that the four through-wall flaws reported in the H4N4 intersection examined in 2014, which consists of a 28.8 inch window [ lj, Sht. 5] out of -552 inch shroud circumference at this elevation, exist in each -29 inch segment of the shroud circumference. This is considered to be a bounding treatment of the possibility for additional through-wall flaws in the shroud at this elevation. Leakage rates are calculated for one and two cycles of operation. File No.: 1200283.303 Page 14 of21 Revision: I F0306-0IRI

7.0 RESULTS The allowable flaw size for an axially oriented flaw in the HNP I core shroud, considering the lower bound fracture toughness of 50 ksi-in°5 and un-irradiated yield strength (conservative) is: LEFM: 45.78 inches Limit Load: 431.76 inches The 2 year, end of interval, bounding flaw size is 14.05 inches. This flaw bounds aU reported axiaHy oriented indications. The calculated allowable operating interval for the reported indications is approximately 38 years using the bounding IGSCC crack growth rate. No re-inspection interval greater than I0 years is currently allowed in BWRVIP-76 [2]. After one cycle of operation, the core shroud structural margin, on fracture toughness, is 4.71. The required structural margin is 1.39. The cumulative leakage rate for the core shroud for all reported and assumed through-waH flaws is: 1 Cycle of operation: 85.8 GPM 2 Cycles of operation: 127.0GPM This includes the through-wall indication in the V6 weld, for which the structural evaluation is documented in Reference [19].

8.0 CONCLUSION

S The axially oriented indications reported in the HNP 1 core shroud are acceptable as-is for at least one additional cycle of operation. The calculated evaluation 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 pro,~ded in the ASME B&PV Code (Code year and addenda as approved for the current operating interval for the plant) [4] or BWRVTP-76 [2], as appropriate, depending on whether the flaws are located in the base material or adjacent to the core shroud welds. The cumulative leakage rates calculated for all reported and assumed through-wall cracks in the HNPl core shroud, after one and two operating cycles of2 years each, are 85.8 GPM and 127.0 GPM, respectively. These leakage rates should be evaluated against leakage rates assumed in current licensing basis (CLB) analyses. SI recommends that the Southern Nuclear Operating Company (SNC) perform additional inspections of the core shroud at the H4 elevation with the objective to identify extent of through-wall, axially oriented indications and the whether existing flaw growth is within the bounds of analytical assumptions. This would allow SNC to refine the shroud leakage assumptions made in this report. The schedule for future inspections would depend on the available margin and the sensitivity to additional flaw growth beyond the one operating cycle addressed in this report. File No.: 1200283.303 Page 15 of21 Revision: I F03()6.{) I R I

l}Sinrctunrllnlsgr/ly Associates. Inc.*

9.0 REFERENCES

1. H 1R26 Inspection Results:

a. INR H1R26 IVVI-14 Shroud OD Indications observed during VT-3

b. JNR HI R26 IVVI-14 Core Shroud Plate Indications

c. JNR HIR26 IVVI-14 Shroud ID-OD-Vertical Weld 4, Horizontal Weld 4 Intersection

d. JNR HIR26 IWI-14 Shroud ID Indications observed during VT-3

e. JNR HI R26 IVVI-14-12 -Shroud ID Vertical Weld V8 @ H5

f. GE Hitachi Customer Notification Form CNF-SHRD-00 l RO

g. GE Hitachi Customer Notification Form CNF-SHRD-002 R1

h. GE Hitachi Customer Notification Form CNF-SHRD-003 RO

1. GE Hitachi Customer Notification Form CNF-SHRD-004 RO J. GE Hitachi Customer Notification Form CNF-SHRD-005 R I

k. GE Hitachi Customer Notification Form CNF-SHRD-006 RO I. GE Hitachi Customer Notification Form CNF-SHRD-007 RO

m. GE Hitachi Customer Notification Form CNF-SHRD-008 RO

2. BWRVJP-76, Revision 1: BWR Vessel and Internals Project, BWR Core Shroud Inspection and Flaw Evaluation Guidelines. EPRI, Palo Alto, CA: 2011. 1022843.

3. BWRV/P-100-A: BWR Vessel and Internals Project, Updated Assessment of the fl*acture Toughness of Irradiated Stainless Steel forB WR Cnre Shrouds. EPRI, Palo Alto, CA: 2006.

1013396.

4. ASME Boiler & Pressure Vessel Code

a. Section II Part D, 200 I Edition.

b. Section XI, 20 10 Ed.

5. Transware Report No. SNC-HA1-002-R-001, Revision 0, ..Edwin I. Hatch Unit 1 Fluence Evaluation at End of Cycle 25 and 49.3 EFPY," SI File No. 1200283.201.

6. Inspection Uncertainty:

a. TR-105696-R/5 (BWRVIP-03) Revision 15: BWR Vessel and Internal<> Project, Reactor Pressure Vessel and Imernals l:."xamination Guidelines. EPRI. Palo Alto, CA: 2011.

1025142.

b. BWRVIP Letter 2014-015, New NDE Demonstrations, February 3, 2014, SI File No.

1200283.207. File No.: 1200283.303 Page 16 of21 Revision: 1 F0306-0IRI

l)StntctunriiRflgrtty Associates, Inc.*

7. BWRVIP-14-A: BWR Vessel and Intemals Project, Evaluation o.fCrack Growth in BWR Stainless Steel RPV Internals. EPRI, Palo Alto, CA: 2008. 1016569.

8. BWRVIP Letter 2012-074 from Chuck Wirtz and Randy Stark to All BWRVIP Committee Members, "Superseded "Needed" Guidance Regarding Crack Growth Assumptions," SI File No.

1200283.205.

9. Tada, H., Paris, P., Irwin, G., The Stress Analysis of Cracks Handbook, 3rd. Ed., ASME Press.

2000.

10. T. H. Okiishi, B. R. Munson, and D. F. Young, "Fundamentals of Fluid Mechanics," John Wiley

          & Son, Fourth Edition, December 13, 2001.

II. Hatch Nuclear Plant Unit I Core Shroud Drawings:

a. Southern Nuclear Company Drawing No. 1-BN-6-6, Revision 1, "Core Shroud Weld Section and Details," SI File No. HTCH-02Q-205.

b. Southern Nuclear Company Drawing No. 1-BN-6-5, Revision 2, "(Inside View} Core Shroud Weld Identification Layout," SI File No. HTCH-02Q-205.

12. Design Input Request, Revision 1, Sl File No. 1200283.200.

13. Southern Nuclear Company Nuclear Management Guideline, NMP-CH-005-GL01, Version 5.0, "E. I. Hatch Water Chemistry Strategic Plan," April2012, SI File No. 1200283.206.

14. BWRVJP-62 Revision}: BWR Vessel and lntemals Project. Technical Basis for In~pection Relief for BWR lntema/ Components with Hydrogen Injection. EPRI, Palo Alto, CA: 2011. 1022844.

15. NIST Chemistry WebBook, http://webbook.nist.gov/chemistry!lluid/.

16. Rahman S. et. a/, "Crack-Opening-Area Analyses for Circumferential Through-Wall Cracks in Pipes -Part 1: Analytical Models," International Journal of Pressure Vessels and Piping, Volume 75, Pages 357-373, 1998.

17. Rahman S. et. al, "Crack-Opening-Area Analyses for Circumferential Through-Wall Cracks in Pipes- Part lll: Off Center Cracks, Restraint of Bending, Thickness Transition and Weld Residual Stresses," International Journal of Pressure Vessels and Piping, Volume 75, Pages 397-415, 1998.

18. Mathcad Version 14.0 M035 (14.0.3.374), Parametric Technology Corporation.

19. SI Calculation Package 1200283.302, Rev. I, "Hatch Nuclear Power Plant Unit 1 Core Shroud Vertical Welds V5N6 Crack Growth and Fracture Mechanics Evaluation".

File No.: 1200283.303 Page 17 of21 Revision: I F0306*01RI

estntctumllllfsgrlly Assoc/al9s, Ina!' rt C!ft 1-D-a-e'J l1P II' SMIWII 111:&11 F'1.JIIIa I TIP 17 t.CTIVI: nn.

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                                                                                                                  ~187tt nAI I-117B4 C - UIIPiliiUIDG L IR1JID: Cl1 I s-t'Mf>B c - . EUC1111CI                                                                                                            -- 1-BH-5 Figure I. HNP1 Core Shroud Configuration [12)

File No.: 1200283.303 Page 18 of21 Revision: I F0306-0IRI

l)structurallntsgrtty Assoclatfls, Inc.* 100 1-n zoe eto m eJO zoo ~a e6o 17D 2no eo;o JOO 31D m 3:14 ~~ l~ o 10 eo 30 40 !10 60 7Q SD 90 too 110120 130 140 !:SO 160 170 180 I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I 1I I I I I I H'l CO Jlj 70 00 90 IDD 110 120 1::11) 1*0 "0 11'.0 17C 100 I I I I I I I I I I I I I I

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                                                          --               S-1~ ~- SMlPP.ltL.DII<G ~ CR"t1IIX<< CO >

S*lti149 nQU S*IB7'!14 CS\/Ji S><IPJUIUING' IIRTlllDt Cll I S*l~ t!ZNU!Ol. EUt:TltiCl Figure 2. Previous HNPl Core Shroud Inspection Results (12) File No.: 1200283.303 Page 19 of21 Revision: 1 F0306-0IRI

l)sttuctumllntsgrtty Associalrls,lnc.*

1fl F(.\) ~,(I+ 1.2$.\ l 0< .A :0:: I

                  -Ciadt Opening Area:

_G(.\)

                                                 .
                                              =)." + ,625).
                                                           ~

Figure 3. LEFM Solution for a Single Through-wall Axial Crack in an Internally Pressurized Cylinder (9) File No.: 1200283.303 Page 20 of2l Revision: 1 FOJ06-0IRI

Figure 4. LEFM Solution for an Inrmite Array of Through-wall Cracks in a Plate with a Membrane Load [9) File No.: 1200283.303 Page 21 of21 Revision: 1 1'0306-0IRI

eBtntc:lui811Rt8grlly Associates, lna.0 APPENDIX A CALCULATIONS File No.: 1200283.303 Page A-1 of A-14 Revision: 1 F0306-0IRI

e81ntclurllllllf8grlty Associat9s, Inc.* Responsible Engineer: Daniel Sommerville Date: 17FEB2014 Plant HNP1 Calculation Title: H4 axially oriented weld Haw evaluation Objectives: Perform:

1. Flaw evaluation and leakage calculations for axially oriented flaws in the HNP1 core shroud

2. Leakage calculation for the through-wall indications.

Methods: 1. Using methods generally consistent with BWRVIP-76

2. Use handbook LEFM solutions to address:

a LEFM of axially oriented crack in a cylinder

b. Crack opening area of axial crack in a cylinder

c. Plastic zone size

3. Use Bernoulli flow solution to calculate leakage flow. see method given in BWRVIP-76 with an additional loss factor of 0.63 to account for entrance and exit of contraction. No loss is considered for crack path.

4. Bounding SCC CGR of 5E-5 in/hr is used. consistent with BWRVIP-14-A and BWRVIP-76. Rev 1.

5. Allowable static initiation fracture toughness. KIC. equal to 50 ksi-in0.5 is used as given in BWRVIP-100.

Inputs: 1-" I .)-

Ri :=--*m Inside radius of shroud at H4 location 2

t := 1.5-in Wall thickness of shroud at H4 location RIPDab := 11.6*psi Level NB RIPD RIPDcd := 295*psi Level CIO RIPD SFab := 2.77 LeveiNB SF SFcd := 1.39 Level C/0 SF

              -   -5 in         sec length crack growth rate CGR := )*10      *-

hr

   <1)' := 18900-psi            Yield strength of shroud material. unirradiated. 550 F.

m := -'1200-psi Flow stress of shroud material, unirradiated. 550 F. 6 E := 25.55 x 10 -psi Elastic modulus. 550 F l' := 03 Poisson's ratio File No.: 1200283.303 Page A-2 of A-14 Revision: I F0306-0IRI

  ~Sinn:lurallntegrlty Associates, Inc.*

KIC := 50000-psi*in0.5 Fracture toughness. for fluence > 3E21 n!cm2. u := 12*in Initial flaw size l_a := ~S.785in Allowable flaw size. Calculated by iteration LEF := O.U9*in Length evaluation factor gc:="~86.6*lbm -

in

                         -                   Newton's proportionality constant lbf "I s*

Ibm p := 62.J- Density of water at 50 F ft3 Calculations:

1. Calculate allowable operating interval.

Rm:=Ri+.! Rm = 88-in 2 (Rm) u := RIPDcd*-- t 1_a 2

\:=--- ).. = 1.993 (Rm*t) 0.5 "1. 0.5 F(:\) := { 1 + 115* x.. l if o~ " ~ 1 F(:\) = 2393

((0.6 + 0.9*:\)) if 1 ~ ).. ~ 5 ry = 1.11~-in Kl(:\) := SFcd*rr* [ot*l"1~a + ry

                                             )~OJ
                                              ~ *F(:\)            Kl(:\) =    ~9999*psi *in°* 5 1 a - 1 i- 2* LEF 1 1 yr Inter..-al :=                         * -* -*.:.....          lnter..-al = 3S.227*)'T 2*CGR           2~ 365 hr File No.: 1200283.303                                                                            Page A-3 of A-14 Revision: 1 F0306-01RI

eSinlc:tunr/ lllt8grlly Associates, Inc.* 2 Calculate structural margin after one operating cycle. 1 f := 1 i + 2*LEF + 2*2*,T*CGR*365 day *2~*..!!_ 1f = l~.05*in

    -        -                   *              )T  day  -

l_f

                '\
    >.. :=      -                                       A= 0.611 (Rm*t)0.5
                 .            "10.5 F(A) :=      I + 115*A"'.r 1.1                     if  o~A~   1     F(A)  = 1.211

((0.6 + 0.9*A)) if 1 ~ A~ 5 I)' := f-.( KIC)2 I)' = l.ll-'*in

            -*~ 1, cry
                                  )~0.5
                       ,..1 f KI(A) := cr* [ ~*l2 + I)' ~ *F(>..)                  KI(A)  = l060l*psi*in°*5 KIC                                                           !\iargin = 3.393
   !\*largin := KI{>..)                                 Margin = 4.717 SFcd 3 Limit Load allowable flaw size.

lallow := 1.58-{Rm*t)O.;,_

                               -[,..!my_   '\-J0.5 1         tallow=   ~31.762*in
                                   \~ cr /

File No.: 1200283.303 Page A-4 of A-14 Revision: 1 F0306..01RI

  ~Stnn:tunrllllfflgrtty Assocfatss, Inc.*

4. Calculate leakage at end of one operating cycle.

Flaws- 8 0" at V4/H4 @ --15 inches 5.5" at V4/H4 @ --6 inches 6.2" at V4/H4 @ --1 inches 5.3" at V4/H4@ -3 inches 20.3" at V6 8.0 The first four flaws are the V4/H4 indications. 5.5 l_i:= 6.2 *in 5.3 20.3 This flaw is the V6 indication.

                                                                                 "10.05 7.55 1 f := 1 iT 2*LEF

2*2*}T*CGR*365 day *1~-~

    -        -                                    )T     day              1f   =    8.25   *Ul 7.35 22.35 Iterate to find plastic zone size correction factor on a flaw specific basis:

i := 0, 1.. .s Iteration 1: (0.~37'1 t_f 0.329

                 .,
    >..:=--*--                                                            >.. = 0.359 (Rm*t}0.5                                                             0.32 0.973 1.113
                   .              05                                                  1.065 1

F(>..) := It+ 1.25->.. 1. if 0~ >..~ 1 Fj\1 = 1.078 ((0.6 + 0.9*>..)) if 1 ~ >.. ~ 5 1.062 l..Si7 File No.: 1200283.303 Page A-5 of A-14 Revision: I F0306-01Rl

10640 8826 Kl1{A,I_f) := SFcd*cr* [ C2

                               ':t*   f)r.s *F(A)                              KI1(A-,1I -f-)

I = 9332

                                                                                                              .* 0.5
                                                                                                          *pSl*Ul 8681 21058 r

0.05

                                                               ")                                  0.035 tyl(A,l_f) := _l_{Kll(A,1 f)                               ryt(Ai,l_fi)     =  0.039 *in 2*7f           cry 0.034 0.198 Iteration 2:

0.442 I f 2 + ryl(A,l_f) 0.332

                                                                               ). 1fA.. 1 f.)=
    }..2(A,l_f) := - - - - -

(Rm*t)0.5 '\ **-** 0.362 0.99

                                             ")0.5 F2(A,1_f) :=      ( 1 + 1.25*}..2{A,1_f)'"".        if 0 :S Al(>.,l_f) :S 1

((0.6 + 0.9*}..2(>.,1_f))) if 1 :S }..2(A,l_f) :S 5 1.115 1.067 F22{*>..,t

1 -I r.) = 1.079 1.063 1.492 Kl2(A,1_f) := SFcd*cr*[ ':t* ( 2 + ryl(>.,l_f))~0.5 1f

                                                           ~ *F2(>.,t_f) 10714 8876
                                                                                   ,')* 1_f i.) =

J(T'){'

                                                                               .                    9~88
                                                                                                     ~       ..
                                                                                                          *pSI*Ul 0*5 8730 21447 File No.: 1200283.303                                                                                                 Page A-6 of A-14 Revision: I F0306.01Rl

e&lnlctutal llltBgrlly Assaciatfls, Inc." 0.051 1 0.035 ty2()1.,l_f) := _I_.(Kl2(X,l f))- 2*7\ ay cy2("-i'1_fi) = 0.039 *in 0.03-l 0.205 1397 1.1-ll PC()I.,1 D := ry2()1.,1 f)- cyl()l.,l_f) *100

            -              cy1()1.,l_f)

PC("-i,l_fi) = 1.205 1.12-l 3.735 Kll()l.,l_f) := K.l2()1.,1_f) tyl("-,l_f) := cy2()1.,1_f) Iteration 3: 0..!42 1f

                    -=- + ry1()1.,1_f)                                                              0.332 2

h2()1.,1_f) := - - - - - ').'f>-..,1

                                                                                 ., 1 -t r.)=   0.362 (Rm*t)0.5 0.99 F2(>-.,t_f) :=   (1 + 1.25*>-.2{"-.l_f)"'"."'*)0.5  if 0 ~ >-.2()1.,1_£) ~ 1

((0.6 + 0.9*h2()1.,1_f))) if 1 ~ >-.2(>-.,l_f) ~ 5 1.115 1.067 F2(>--i,l_fi) = 1.079 1.063 1.-192 File No.: 1200283.303 Page A-7 of A-14 Revision: I F0306-0IR1

  !S3stnrotum1 IBtJgrlty Associates, Inc:.*

Kll(>.,l_f) := SFcd*cr*[ it* ( 2 + tyl(>.,l_f)Jj)10.5*F2()..,l_f) 1f 10715 8876

                                                                                                       .. 0.5 KI2(>..,1 f.)
                                                                              . l -   l.
                                                                                         =    9388 *pst*m 8730 21462 r

0.051

                                                           .,                                0.035 sy2(>..1_f)    == _1 _(KI2(>..1 f)                 sy2(>.i,1_fi)   =   0.039 *in 2*it \. cry 0.034 0.205 0.019 0.013 PC(>. 1 f) := sy2(X.,l f)- ryl(>.,l f) *100 PC(>..,l r.) =    0.014 I-                  1(>. 1 f)                                     l -   l.

ry *- 0.013 0.136 ry(>.,t_f) := tyl(>..,1_f) 0.-Ul 1f

                  -=- + rv(>. 1 f)
                   ')     ~     I-0 _,,_
                                                                                            ~~,

Xf(>.,1_f) := - - - - - - (Rm*t)0j 0.991 File No.: 1200283.303 Page A-8 of A-14 Revision: 1 F0306-0IRI

eSinlctunrii/Qgrlty Associates, Inc.* G()..,l_f) := M()..,l_f)~ + 0.625-M()..,l_f).l if 0 5 M()..,l_f) 5 l

                                              ~                 .

O.U + 036-M()..,l_ff + 0.72-M()..,l_f)" + OA05*M()..,l_f)

if 1 5 M()..,1_f) ~ 5 0.219 0.118 r.l )...,1

    -1. I - I f-1 =    0.1.$2 0.111 1.583

( 0.011 3 16.009 X 10-IJ COA().. l f) := ---2-;t*Rm-t-G().. 1 f) COA( )..i,l_fi) = 7.268 x 10-3

               ,_            E                  *-
                                                                                                    -3
                                                                                       !\.5.675 X 10

1 0.081 r l..$56 0.781 2

Q()..,1_f) := 0.63COA()..,1_f)*J -gc-~Ikd n!)... .l

                                                                   "<\ I' -

f.)= I 0.9.l5 . gal mm 0.738 10.523,/ Assuming equal flaw density in all 12 high fluence regions + 1 through-wall flaw in VG gives: 3 gal Qtotal := 12-I Q{.\*l_fil + Q()...l,1_f.l) Qtotal = 57.6--_ mm i= 0 Assuming an equal distribution of identical leaking flaws in the core shroud at the H4 weld + 1 through-wall flaw in VG gives: 3

L Q("i*ui_l*2;t-Rm i= 0 gal 28.8-in + Q(."4'1_f.t) = 85.8* min File No.: 1200283.303 Page A-9 of A-14 Revision: I F0306-0IRI

  !S)Structu~allnf8grtty Assoclatfls, Inc.*

5. Calculate leakage at end of two operating cycles.

Flaws- 8.0" at V4/H4 @ --15 inches 5.5" at V4/H4 @ --6 inches 6.2" at V4/H4 @ --1 inches 5.3" at V4/H4 @ -3 inches 20.3" at V6 S.O The first four flaws are the V4/H4 indications. 5.5 1_i := 6.2 *in 5.3 20.3 This flaw is the V6 indication. 11.802 9.302 1 f := 1 i + 2-LEF + 2*-l*)T*CGR-365 day *2-l-~

     -       -                                       )T  day             l_f =     10.002 *in 9.102 2U02 Iterate to find plastic zone size correction factor on a flaw specific basis:

i := 0,1..-l Iteration 1: l_f "I ll(~::~~\ A:=--*-- A= o..m 05 (Rm*t) 0.396 l.O-l9

                                                                                    '1.153 l.09S f()..) :=     (.1 +        "10.5 1.15-}... if 0 :s; ),. :s; 1                 Fj)..i)"'    U12

((0.6 + 0.9*>-)) if 1 5 ),. 5 5 1.09-l 1.5U File No.: 1200283.303 Page A-10 of A-14 Revision: 1 F0306-0IRI

l)Stntr:trnallllf8gr/ly Assocla$s, Inc:." 11944 10094 f)~0.5 Kll(~,l_f) := SFcd*u* [ Of* ( 2~ 1

                                              *F(~)                       Kll(~-.1 1 - 1 f-)* =   10604 *psi*in0.5 9948 22854 0.064
                                                           ,                                    0.045 ryl(~.l_f) := _1 *(Kil(~.l           f))-               rv1(~-.l     r.)   =   0.05 *in 2-';t          cry

I - 1 0.044

                                                                                                   , ....

0-~~ Iteration 2: 0.519 1f

                  -:; + ryl(~.l_f)                                                             0.409
    "-2(~.1 f ) : = - - - - - -                                           ),.')f~** 1
                                                                          " \ I -1, f.)=      0.44
           -          (Rm*t) 05 0.4 1.069
                                           .,)*0.5 F2(~.1_f) :=    (1 +  l.25*"-2{~.1_f)"'"         if 0 ~ "-2(~.1_f) ~ 1

((0.6 + 0.9*"-2{~.l_f))) if 1 ~ "-2(~.l_f) ~ 5 1.156 1.099 F2f~.* 1

                                                                            \    1 r.i =
                                                                                    - t,l 1.114 1.095 1.562 KI2(~,1_f) ;=  SFcd*u*  [ (I2 +

Of* f ryl(~.l_f) ~

                                                       )~0.5
                                                             -F2(~.1_f) 12040 10160 0
                                                                                                          .. *5
                                                                                '\ 1_ f i,)

Kll( "i* = 10678 *pst*m 10012 23346 File No.: 1200283.303 Page A~ll of A~l4 Revision: 1 F0306-0IRI

  !S}Bintctul1lllnfegrlty Assaciatss, Ina."

0.065

                                                             ,                                   0.046 tyl()l.,l_f) := _1__ (KU()I.,l      f))~                rv2()1..,1  r.)t = 0.051 *in 2*or          ay                        ~   . 1 -

0.045 0.2J3 1.62 1313 PC()I.,l f) := ty2()1.,1_f)- tyl()l.,1_f) *100

              -              tyl()l.,1_f)

PC()I.i,l_fi) = 1391 1291 4352 Kll{)l.,l_f) := KU()I.,l_f) tyl{)l.,l_f) := ty2()1.,1_f) Iteration 3: 0.519 1f

-=-

2

                          + rvl()l.
                             ~

1 f)

'-

0.409

    )1,2()1.,1 f) := - - - - -                                                  ~ 'f>..,t
                                                                                .  , 1 -1. r.) = o.44
              -          (Rm*t)0.5 0.4 1.07 F2()1.,l_f) :=     (1 + 1.25*)\,2()\.,l_f)"",)0.5  if 0 ~ )\,2()\.,l_f) ~ 1

((0.6 + 0.9*)\,2()\.,l_f))) if 1 ~ )\,2()\.,l_f) ~ 5 1.156 1.1 F2()1.i't_ri) = 1.114 1.095 1.563 File No.: 1200283.303 Page A-12 of A-14 Revision: I F0306-01Rl

eSitur:tumlllliJgrlty Assoalalr1s. lnc.8 KI2(>..,1_f) := SFcd*a* [ or* ( l 1f )~0.5*F2(>..,1_f)

                                           + ryl(>..,1_f) ~

12042 10160

                                                                              "'\ i' 1_f i)

Tl'T1 f),.

                                                                                                 =    10679 *pst*m
                                                                                                               . . Oj 10013 23368 0.065
                                                                .,                                   0.046 n ~ _1 *(KI2(>..,1_f) \-

ry_"('"* 1_.., .- ) ry2(>..i,1_ri) = o.051 *in 2*or cry 0.045 0.243 0.026 0.017 PC(>..,l f) := ty2(>..,1 f)- ryl(>..,l f) *lOO pr-1>,..,1 f.)I. = 0.019

              -               ry1(>..,1_f)                                    -\    I -

0.017 0.184 ry(>..,1_f) := ry2(>..,1_f) 0.519 1f

                    -:; + ry(>..,l_f)                                                             0.409 M"(>-.,1  n == - - - -                                              M"(>...,l
                                                                                , I- I f.) =     0.44
            -           (Rm*t) .5 0.4 1.07 File No.: 1200283.303                                                                                                  Page A-13 of A-14 Revision: 1 F0306.01Rl

G(>-.,1_f) := Af(>-.,1_f)~ + 0.625-Af(>-.,1_f)'~ if 0 $ Af(>-.,1_f) ~ 1

                                                  ..                 *                    .1 OJ..t + 0.36-Af(>-.,1_ff.,. 0.72-Af(>-.,1_f) + 0A05-Af(>-.,1_f)       if 1 ~ Af(>-.,1_f) ~ 5 0.31:5 0.185
    "'I >-.-,1 r.l   1 -I f-J =    0.217 O.li6 1.965

( 0.016 I 9.-l3i X 10- 3

(j COA(>-.,1 f) := ---2-":"I"*Rm-t-G(>-.,1 f) E - COA{\,1_fi., =I 0.011 l 8.998 X 10-3

                                                                                           '-      OJ

( 2.09:5

                                                                                           .,, ..

1_, Q(>-.,l_f) := 0.63 COA(>-.,1_f)*J 2-gc-RIPDcd Ql\*l_fj) = l...J.ll . ~ p 1.17

                                                                                      ,13.068 Assuming equal flaw density in all12 high fluence regions+ 1 through-wall flaw in V6 gives:

3 Qtotal := 12* I Q(\,l_fj) + Ql >-...t,l_f .tl Qtotal = 8-U* gal min i= 0 Assuming an equal distribution of identical leaking flaws in the core shroud at the H4 weld + 1 through-wall flaw in V6 gives: 3 I Ql\*l_fiJ*h*Rm i= 0 gal 2S. *in + Q(_ >-.4 ,t_f.i) = 127* min 8 File No.: 1200283.303 Page A-14 of A-14 Revision: I F0306.01Rl}}

NL-14-0840, Lnservice Inspection Program Owner'S Activity Report for Outage 1R26: Difference between revisions

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9.0 REFERENCES

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NL-14-0840, Lnservice Inspection Program Owner'S Activity Report for Outage 1R26: Difference between revisions

Revision as of 04:45, 4 November 2019

Text

Enclosures:

Background

1.0 INTRODUCTION

8.0 CONCLUSION

9.0 REFERENCES

1.0 INTRODUCTION

8.0 CONCLUSION

9.0 REFERENCES

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