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CNS Inservice Insp Summary Rept Fall 1998 Refueling Outage (RFO-18)
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Issue date:	03/15/1999
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COOPER NUCLEAR STATION INSERVICE INSPECTION SUMMAR.Y REPORT FALL 1998 REFUELING OUTAGE (RFO-18) 1 DR 0 K 05 0 8

TABLE OF CONTENTS l

l TABLE OF CONTENTS . . .. . .. . . . . . . . . . . . . . .. . . . . , . .i l !

I. NIS-1 Form... . . . . . . . . . . . .. . . . . . . . . .... . . . . . . . .1 II. Inservice Inspection Summanf Report... . . . . . . . . . . . . .2 l 1 i

l.0 Introduction.. . . . . . . . . . . . . . . . . . .. .. . . . . . .. . ....2 2.0 Summary of Examinations .. . ... .. . . . . . . . . . . . . . . .. ......2 2.1 Review of Examination Results . .. .. . . . . . .. . . . . . . . .2 j l Table 1 - Summary of Examinations .. . .. . . . . . . . . .4

! 3.0 Relief Request and Regulatory Correspondence .. .. . .. . .10 Table 2 - List of Relief Requests and Regulatory Correspondence .... . . .10 l

4.0 Repairs and Replacements ..... . ... .. . . . . . . .12

! Form NIS-2 Owner's Report for Repairs or Replacements . . .. .. ..... ..... 13 l Table 3 - Summary of Repairs or Replacements.. . .. . . .14 i

5.0 ATTACHMENTS ATTACHMENT 1 Abstract of Examination Attachment 1.1 Examinations Performed By GE,23 Pages Attachment 1.2 Examinations Performed by Westinghouse,-3 Pages Attachment 1.3 Examinations Performed by NPPD,2 Pages 1

ATTACHMENT 2 Examination Personnel Attachment 2.1 GE Personnel,1 Page l Attachment 2.2 Westinghouse Personnel,1 Page l

Attachment 2.3 NPPD Personnel,4 Page i

ATTACHM5'.NT 3 Examination Procedures ,

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Attachment ..I GE Procedures,1 Page l Attachment 3.2 Westinghouse Procedures,1 Page Attachment 3.3 NPPD Procedures,1 Pages Attachment 4 Equipment and Materials ,

l Attachment 4.1 GE Equipment and Materials,3 Pages

! Attachment 4.2 Westinghouse Equipment and Materials,1 Page Attachment 4.3 NPPD Equipment and Materials,1 Page L

- Attachment 5 Examination Results Accepted By Analytical Evaluation I l

CNS RPV Flaw Evaluation Handbook-Calculation No. NEDC98-048 l

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n FORM NIS-1 OWNER'S REPORT FOR INSERVICE INSPECTIONS l As required by the Provisions of the ASME Code Rules

1. Owner: Nebraska Public Power District,

2. Plant: Cooper Nuclear Station, PO Box 98, Browmille, NE 68321

3. Plant Unit: One 4. Owner Certificate of Authorization (if required): N/A

5. Commercial Service Date: J.pjy,1974 6. National Board Number for Unit: 20762

7. Components Inspected: See Attached Report Summary R. Examination Dates: 5/20/97 (End of Last Outage ) to 12/20/98

9. Inspection Period Identification: First Period - 3/l/96 to 6/30/99

10. Inspection Interval Identification: Third Interval- 3/1/% to 2/28/06

11. Applicable Edition of Section XI: (ISO 1989 Edition, No Addenda. OWE) 1992 Edition,1992 Addenda

12. Date/ Revision ofInspection Plan: OSI) Revision 1.1, November 3,1998, GWE) Revision 0, February 24,1998

13. Abstract of Examinations and Tests: See Attached Report Summary.

14. Abstract of Results of Examinations and Tests: See Attached Report Summary

15. Abstract of Corrective Measures: See Attached Report Summary CERTIFICATE OF COMPLIANCE We certify that a) the statements made in this repr* are correct, b) the examinations and tests meet the Inspection Plan as required by the ASME Code,Section XI, and c) corrective measuri i taken conform to the rules of the ASME Code,Section XI.

Certificate of Authorization No. N/A Expiration Date N/A Signed - AflA . as Date 3/IS!%

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g CERTIFICATE OF INSERVICE INSPECTION I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and the State of Nchraska and Employed by llartford Steam Boiler Insnection and Insurance Comnany of Hartford. Connecticut have inspected the components desenbed in this Owner's Report during the period May 20,1997 to December 20,1998, and state that to the best of my knowledge and belief, the Owner has performed examinations and taken corrective measures described in this Owner's Report in accordance with the requirements of the ASME Code,Section XI.

By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the examinations and corrective measures described in this Owners's Repat. 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 from or connected with this inspection.

/ Commissions //O N 1 Al l brector's Signature Nahonal Board, State, Province, and Endorsements Date 44 19 N l

INSER VICE INSPECTION SUMMA R Y REPOR T NPPD Cooper Nuclear Station Unit I Refueling Outage RFO-18 l

1.0 INTRODUCTION

Nondestructive examinations (NDE) were performed during the Fall 1998 Refueling Outage (RFO-18) at Nebraska Public Power District's (the District) Cooper Nuclear Station (CNS), to ensure the current and future structural integrity of applicable safety related systems and components.

Examinations meet the requirements of:

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= ASME Section XI 1989 Edition j

= l ASME Section XI 1992 Edition with 1992 Addenda (Subsection IWE).

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The Third Ten-Year Interval Inservice Inspection (ISI) Program For Cooper Nuclear Station, Revision 1.1 )

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. The First Ten-Year smernal Containment Inspection Program for Cooper Nuclear Station, Revision 0.

RFO-18 is the second refueling outage in Period 1(3/1/96 to 6/30/99) ofISI Inspection Interval 3 (3/1/96 to 2/28/06). This Summary Report submittal is required by ASME Section XI, IWA-6000 and includes inspections and tests performed between May 20,1997 and December 20,1998 (Operating Cycle 18).

RFO-18 is the first outage the Reactor Pressure Vessel (RPV) shell weld inspections and Containment component inspections have been performed at CNS.

2.0

SUMMARY

OF EXAMINATIONS Nondestructive examinations and tests were performed using Visual Testing (VT), Liquid Penetrant l (PT), Magnetic Particle (MT), and Ultrasonic (UT) inspection techniques. A total of 266 I components were examined / tested and nine (9) pressure tests were completed during Operating Cycle

18. In several cases, one component may refer to a group of examinations performed for a specific requirement.

Table 1 " Summary ofExaminations" lists components examined for ASME Section XI or augmented requirements and provides a limited description of the indications and/or conditions which were recorded. Additional details summarizing individual component examinations are provided in " Abstract of Examination."

2.1 Review of Examination Results l

Of high interest to CNS is the overall condition (health) of the RPV shell welds, nozzle to shell welds, and intemal components. CNS actively participates in the BWR Vessel Internal Project (BWRVIP)

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and has performed augmented examinations of the RPV Internals in addition to RPV examinations required in accordance with ASME Section XI (the Code) and 10CFR50.55a(g)6(ii)(A).

Examination results identified flaw indications in two (2) RPV shell welds and one Main Steam Nozzle to vessel weld (Table 2, Items 15,16,18,22). Additional examinations verified previously identified indications in a Feedwater Nozzle to vessel weld (Table 2, Items 1, 2) and the internal Core Spray piping welds (Table 2, Items 3,4,17,21). All of the indications have been evaluated using !

analytical evaluation methods of the Code Subarticle IWB-3600 and found to be acceptable for continued operation.

RPV Shell Weld Inspection Automated ultrasonic examinations of the RPV Shell longitudinal and circumferential welds were completed during RFO-18 from the RPV inside diameter using the GERIS 2000 ID Szanner and ,

supplemented by manual UT exams to maximize coverage. Five (5) flaw indications in longitudinal !

welds VLA-BA-3 and VLC-BB-2 were identified in the RPV shell welds which did not meet the i acceptance criteria of the Code, Subarticle IWB-3500 and were accepted by analytical evaluation methods of the Code, Subarticle IWB-3600.

The following discussion supports the conclusion that the indications are original fabrication related discontinuities characteristic of slag deposits and are not service induced.

All five (5) indications are attributed to slag deposits, which based on industry experience, has remained as a result ofinsufficient back gouging and/or cleaning of the weld root area during the fabrication process. i

. The indications were independently plotted and evaluated to be in the mid wall region of the weld root. The vertical seam welds in this 218 inch diameter Combustion Engineering RPV are of a double "U-groove" type weld configuration. ,

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. The nature of these indications is siniilar to indications seen by General Electric in other BWR RPV assembly weld examinations.

. The construction radiographs were reviewed; however the indications could not be identified. When reviewing the vessel radiographs, indications from slag inclusion in thick walled vessels are not always apparent since detection oflack of thin film slag deposits is dependent on the alignment of the radiograph beam parallel to the flaw through wall axis. However, these indications inherently have suflicient reflectivity to be characterized by ultrasonic examinations.

Indications of this nature have also been seen in the Main Steam nozzle to weld N3 A and Feedwater nozzle to weld N4D.

The analytical evaluation of the five RPV Beltline indications is provided in Attachment 5 in accordance with the requirements of the Code, Paragraph IWB-3134..

Conclusions Indications in the RPV shell welds, Main Steam nozzle to vessel weld and Feedwater nozzle to vessel welds are attributed to original fabrication related discouinuities. There is no apparent flaw growth Page 3 of 21

mechanism associated with these indications. Based on a summation of the examination results, there are no service induced defects in the Cooper Nuclear Station RPV and continued operation is justified. In summary, all ISI and augmented examinations performed during Operating Cycie 18 meet the requirements of ASME Section XIinclusive of reliefrequests previously submitted.

l TABLE 1

SUMMARY

OF EXAMINATIONS COMPONENTS EXAMINED FOR ASME SECTIO'N XI CREDIT ASME NO. DESCRIPTION CONDITIONS NOTED CAT. COMP.

B-A 17 Reactor Pressure 1) 17 Components required per ISI Program Plan. Exams performed using the Vessel Welds GERIS 2000 ID System and supplemental manual examinations. All required Vertical and Circumferential welds were examined to the extent practical with limitations as described below.

2) 8 Components >90% coverage.

3) 2 Componentsinaccessible 0% coverage.

4) 7 Components < 90% coverage.

5) 1 1 Components contain acceptable flaw indications per IWB-3500

6) 2 components no recordable indications (NRI)

7) 2 Components contain a total of 5 unacceptable flaw indications per IWB-3500 which were excepted by evahiation in accordance with IWB-3600. (VLA-BA-3 and VLC-BB-2). NPPD Calculation No. NEDC 98-048 is provided in Attachment 5.

8) 7 exams include manual pickups.

9) 2 Components examined manually due to access limitations foi the GERIS l

2000 system (NRI).

B-D 18 Reactor Pressure 16 Components (NRI)

Vessel Nozzle Welds 1 Component NVE-BD-N4D, One (1) presiously recorded indication acceptable per IWA-3600 in 1991 and 1995 (Table 2, item 1,2) 1 Component NVE-BD-N3 A, One (1) indication acceptable per IWA-3600 Ref. NEDC98-048 (Table 2, Item 15,16,18,22) l B-F 5 Pressure Retaining 5 Components examined by Smart 2000 system with l l Dissimilar Metal non-relevant indications recorded 1 Welds j i

i B-G-1 36 Pressure Retaining 36 Components Satisfactory (SAT) J Bolting 2" and Greater j in Dia. j i

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l TABLE 1

SUMMARY

OF EXAMINATIONS COMPONENTS EXAMINED FOR ASME SECTION XI CREDIT ASME NO. DESCRIPTION CONDITIONS NOTED CAT. COMP.

B-G-2 13 Pressure Retaining Replaced bolting during drive maintenance. I 3 ISI on existing bolting,13 PSI on new Bolting,2" and Less in bolting.

Dia.

B-J 10 Pressure Retaining 2 Components examined manually (NRI)

Welds in Class 1 8 Components examined by Smart 2000 system with non-relevant indications Piping recorded B-M-2 3 Class 1 Valve Bodies 3 Components (NRI) i l

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TABLE 1

SUMMARY

OF EXAMINATIONS COMPONENTS EXAMINED FOR ASME SECTION XI CREDIT ASME NO. DESCRIPTION CONDITIONS NOTED CAT. COMP.

B-K-1 1 Class 1 Integral I Component (NRI)

Attachments B-N-1 5 Interior Of Reactor Scope of exams includes Jet Pumps 1-10, RPV Interior Wall 0 - 180 Deg., Core Spray Vessel Internal Piping, Core Sprn Spragers and Sparger Brackets. See comp;ementary examinations listed as Augnented.

3 Components (NRI) 1 Component at 2 weld locations (Al & A21) - Core Spray Intemal Piping evaluated with UT. (Table 2, Items 3,4,17,21) See Augmented summary for details.

1 Component (SAT). Core Spray Sparger nozzle indication verified as unchanged since 1995.

B-N-2 6 Integrally Welded Core Scope of exams includes Shroud Support Plate and Gusset Plates and asscx:iated Support Structures and Attachment Welds 0 - 180 deg, Riser Brackets and Welds on Jet Pumps 1-10, CRD Jeerior Attachments to Guide Tubes and Fuel Support Castings at 8 locations.

l RPV l 6 Components (NRI)

C-C 0 Class 2 Integral 2 Components removed from work scope Attachments i C-F-2 1 Pressure Retaining 1 Component (NRI)

Weldsin Class 2 Piping E-A 1 Class MC 1 Component (SAT) Drywell and Suppression Pool (Torus) Metal Containment, VT General.

E-G 2 Pressure Retaining 2 Bolts and Nuts (SAT) PSI 20 Bolts and Nuts (SAT) PSI Bolting Page 6 of 21

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TABLE 1 '

SUMMARY

OF EXAMINATIONS 1

COMPONENTS EXAMINED FOR AUGMENTED INSPECTIONS ASME NO. DESCRIPTION CONDITIONS NOTED CAT. COMP.

INVES 7 NUREG CR3052 Jet Pump Beams 1-10 (NRI)

SIL 465 S1 Nozzle and Mixer Inlet on Jet Pumps 1-10 (NRI)

SIL 420 Sensing Line bracket welds and sensing Lines on Jet Pumps 1-10 (NRI)

SIL 574 Adjusting Screw tack welds on Jet Pumps 1-10 (NRI)

SIL 462 R1S3 RPV Shroud Support Access llole Cover to Shroud Support Plate, two welds (NRI)

SIL 605R1 Thermal Sleeve to riser welds -10 risers (NIU) l RICSIL 078 Adiusting Screw Gaps on Jet Pumps 1-10 (NIU) I INVES 32 BWRVIP-18 Core Spray Piping - 32 Automated UT plus supplemental VT-1 where UT access was Core limited (30 NRI). Existing indications on Al and A21 examined and accepted by l Spray analytical evaluation. See NRC Correspondence (Table 2, Items 3,4,17,21 ) {

l E-A 8 Containment Surfaces 5 Areas of Drywell liner examined Pre & Post 2 Arcu of Torus examined 1 Area of I Augm. Torus examined in 1997 {

GL88-01 10 UT Exams for IGSCC 10 Components examined for ASME Sec.X1 as B-F or B.J Category, no IGSCC detected GL-Plant 2 UT Exams 2 Components examined for plant tracking, IGSCC. No IGSCC detected l Tracking )

NR0619 12 UT Exams of RPV 2 Nozzle Inner Radius examined for ASME Sec XI Nozzle inner Radius 2 Nozzle Inner Radius examined for FR)REGO619 and Bore Regions 2 Zone 5 examined for ASME Sec.X1 2 Zone 5 exams not required 4 Nozzle Bore regions examined fer NUREG0619 REC 75 UT exams for N1 SCC 1 indication removed.

7 Components with recorded geometry 67 Components (NRI)

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TABLEi

SUMMARY

OF EXAMINATIONS COMPONENTS EXAMINED FOR SAMPLE EXPANSION REQUIREMENTS ASME NO. DESCRIPTION CONDITIONS NOTED CAT. COMP.

B-D EXP 2 B-D Category Scope 2 Components (NRI) Reference RI-27 (Table 2, Item 22)

Expansion 4

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I TABLE 1

SUMMARY

OF EXAMINATIONS PRESSURE TESTING ASME NO. DESCRIPTION CONDITIONS NOTED CAT. COMP.

B-P 1 ASME Class 1 1 Component (SAT) Leaks identified at mechanical joints and packing resolved.

C-II Leakage Test -

Systems NB and CRD B-P 1 ASME Class 1 1 Component Alternate Test Performed (SAT). Performed, higler pressure, Leakage Test per pneumatic pressure test in lieu of static (water) head test with RPV cavity flooded as ReliefRequest PR-04 stated in Relief Request PR-04. ASME Section XI credit for exam is pending approval of ReliefRequest PR-04 Rev.1. (See Table 2, No. 23)

.C-H 5 ASME Class 2 RCIC,IIPCI, RIIR, SDC & CS (SAT)

Pressure Test C-11 1 ASME Class 2 & 3 REC (SAT)

D-B' Pressure Test E-P 1 Pressure Retaining Examination Performed in accordance with the requirements of 10CIR Appendix J Components (SAT).

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3.0 RELIEF REQUESTS AND REGULATORY CORRESPONDENCE Table two (2) provides a list of major correspondences related to the ISI Summary Report.

TABLE 2 RELIEF REQUESTS AND REGULATORY CORRESPONDENCE ITEM DESCRIPTION 1 Letter (NLS950240) to USNRC from J. H. Mueller (NPPD), dated December 16,1995, ,

" Report of Feedwater Nozzle Examination Results and Relief Request" )

2 Letter to G. R. Horn (NPPD) from William D. Beckner (USNRC), dated June 7,1996,

" Relief Request from the ASME Code Successive Examination Requirements for Feedwater Nozzle to Shell Weld (TAC No. M94260)"

3 Letter (No. NLS970088) to USNRC Document Control Desk from P. D. Graham (NPPD) dated May 7,1997, " Inspection of Core Spray Spargers and Piping" 4 Letter to G. R. Horn (NPPD) from J. R. Hall (USNRC) dated May 9,1997, " Cooper Nuclear Station - Evaluation of Core Spray Piping Indications During Refueling Outage 17 (TAC No. M95141)"

5 Letter (No. NLS970132) to USNRL bocument Control Desk from P. D. G dated July 14,1997, " Impact of Core Spray Line Crack Indications" 6 Letter to G. R. Horn (NPPD) from J. W. Clifford (USNRC) dated October 23,1997,

" Evaluation of the Third Ten-Year Interval Inspection Program Plan and Associated Requests for Relief for Cooper Nuclear Station (TAC No. M94000) 7 Letter to USNRC Document Control Desk from J. H. Swailes (NPPD) dated February 24, 1993, " Containment Inspection Program" 8 Letter (No. NLS980020) to USNRC Document Control Desk from G. R. Horn (NPPD) dated April 23,1998, " Cooper Nuclear Station Inservice Inspection Relief Requests" 9 Letter (No. NLS980118) to USNRC Document Control Desk from J. H. Swailes, (NPPD) dated August 6,1998, " Containment Inspection Program Relief Requests" 10 Letter to G. R. Horn (NPPD) from J. R. Hall (USNRC) dated June 30,1998, " Request for l AdditionalInformation Related to Requests for Relief From Certain ASME Code Requirements for Inservice Inspection for Cooper Nuclear Station (TAC No. mal 163)"

l 11 Letter to G. R. Horn (NPPD) from D. L. Wigginton for J. R. Hall (USNRC) dated August l

14,1918," Request for AdditionalInformation Regarding Cooper Nuclear Station Third Interval Inservice Inspection Program Plan Requests for Relief (TAC No. MA2138)"

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l ITEM DESCRIPTION l 1

12 Letter (No. NLS980133) to USNRC Document Control Desk from Alan R. Shiever for John H. Swailes (NPPD) dated August 31,1998," Inspection of Reactor Vessel Shell l

Welds" i

13 Letter (No. NLS980145) to USNRC Document Control Desk from M. F. Peckham for l J. H. Swailes, (NPPD) dated September 21,1998, " Inservice Inspection Relief Requests" ;

14 Letter (No. NLS980163) to USNRC Document Control Desk from J. H. Swailes, I

(NPPD) dated October 2,1998," Inspection of Reactor Vessel Shell Welds - Supplemental j l Information" l

15 Letter (No. NLS980178) to USNRC Document Control Desk from J. H. Swailes (NPPD), I dated October 26,1998, " Main Steam Nozzle Weld Indication Relief Request" !

l 16 Letter (No. NLS980182) to USNRC Document Control Desk from M. F. Peckham for J. !

H. Swailes (NPPD), dated October 30,1998, " Main Steam Nozzle to Shell Weld Fracture Mechanics Evaluation" 17 Letter (No. NLS980181) to USNRC Document Control Desk from J. H. Swailes, (NPPD) dated November 06,1998, " Inspection of Reactor Vessel Internal Core Spray !

Piping" !

J 18 Letter (No. NLS980189) to USNRC Document Control Desk dated November 17,1998, !

" Main Steam Nozzle Weld Indication Revised Relief Request" !

i 19 Letter to G. R. Horn (NPPD) from J. N Hannon (USNRC) dated November 23,1998, !

" Flaw Evaluation of Main Steam Nozzle to Shell Weld: Cooper Nuclear Station (TAC No. MA3891)"

20 Letter to G. R. Horn (NPPD) from J. N. Hannon (USNRC) dated November 23,1998,

" Evaluation of Request for Relief RI-04 for Cooper Nuclear Station (TAC No. i MA3454)" 4 21 Letter to G. R. Hcrn (NPPD) from J. N. Hannon (USNRC) dated November 23,1998,

" Cooper Nuclear Station: Flaw Evaluation of Core Spray Piping (TAC No. MA3965)"

22 Letter to G. R. Horn (NPPD) from J. N. Hannon (USNRC) dated December 7,1998, i

" Cooper Nuclear Station Relief Request RI-27, Revision 1 Regarding Alternate i Examinations of Vessel Nozzles (TAC No. MA3891) 23 Letter (No. NLS980028) to USNRC Document Control Desk dated March 19,1999,

" Inservice Inspection (ISI) Relief Request, PR-04, Revision 1, Regarding Pressure Testing of the RPV Head Flange Seal Leak Detection Line" l

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1 4.0 FORM MIS-2 OWNER *S REPORT FOR REPAIRS OR REPLACEMENTS

, A total of 133 Repair or Replacement Activities were performed in the scope of this summary report. FORM NIS-2 is contained in this Section in addition to information provided in Table 3 which summarizes the Repair or Replacements performed and provides a reference to the associated work package.

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FORM NIS-2 OWNER'S REPORT FOR REPAIRS OR REPLACEMENTS As required by the Provisions of the ASME Code Section XI l

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1. Owner Nebraska Public Power District Date December 20,1998 l

PO Box 98 Brownville. Nebraska 68321 Sheet 1 of 9 I

! 2. Plant Cooner Nuclear Station Unit One PO Box 98 Brownville Nebraska 68321 N/A I RepairOrpntsta,EtrWa- 7etrh,m 3 Work Performed by NPPD Type Code Symbol Stamp N/A Authorization No. N/A i PO Box 98 Bromsville. Nebraska 68321 Expiration Date N/A l

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4. Identification of System _ As shown in the Attached Table 1

5. (a) Applicable Ccnstruction Code As shown in the Attached Table

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(b) Applicable Edition of Section XI Utilized for Repairs or Replacements:1989 Edition & 1992 Edition 1992 Addenda (IWE) !

6. Identification of Components Repaired or Replaced and Replacement Components: Shown on Attached Table

7. Description of Work As shown in the Attached Table

8. Tests Conducted: As Shown in the Attached Table

9. Remarks The following Code Cases listed in the Third Ten-year Interval Program were used for repair and replacements:

N-416-1 CERTIFICATE OF COMPLIANCE We certify that the statements made in this report are correct and these repairs and replacements conform to the rules of the ASME Code,Section XI.

Type Code Symbol Stamp N/A Certificate of Authorization No. N/A Expiration Date N/A Signed - #W.~--$.v$.[a Date 5//IW

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CERTIFICATE OF INSERVICE INSPECTION !

I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and the State of Nebraska and Employed by llartford Steam Boiler Insnection and Insurance Comnany of Ilartford. Connecticut have inspected the components described in this Owner's Report during the period May 20.1997 to December 20.1998. and state that to the best of my knowledge and belief, the Owner has performed examinations and taken corrective measures described in this Owner's Report in accordance with the requirements of the ASME Code,Section XI.

By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concermng the examinations and corrective measures described in this Owners's Report. Furthermore, neither the Inspector nor his empk>yer shall be liable in any manner for any personal injury or property damage or a loss of any kind arising from or connected w ith this inspection.

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. ATTACHMENT 1 ABSTRACT OF EXAMINATION I

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ATTACHMENT 1.1 Gs NUcts.4RsNsRGr COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS is-nov-9s Page 1 of 23 ASME COMPONENT ID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BLK 17 COMPONENTS FOR CATEGORY B-A B-A VCB-BA-2 218" GE. BN-3 RPV-04 11/11/98 Geris GE-UT-700V2 16 Vessel Cire Weld Shell 1 to Shell 2. Geometry and fifteen (15) acceptable flaw indications recorded by Geris GE-UT-70IV2 16 Geris ID system. Coverage was 66.3%

B-A VCB-BB-1 218" GE. BN-3 R 216 10/22/98 OL UT-CNS-300V3RO 16 Vessel Circ Weld Shell 1 to Bottom IIcad. 94% coverage achieved. No Recordable Indicators (NRI). 45S UT-CNS-300V3 RO 16 60S UT-CNS-300V3 RO 16 OL UT-CNS-300V3 RO 15 45S UT-CNS-300V' ' ' ) 15 60S UT-CNS-300V3 hu 15 B-A VCB-BB-3 218" GE. BN-3 RPV-08 11/12/98 Geris GE-UT-700V2 16 Vessel Cire Weld Shell 2 to Shell 3. Geometry and forty four (44) acceptable flaw indications Gens GE-UT-70lV2 16 recorded by Geris ID system. Coverage was 81.8%

B-A VCB-BB-4 218" GE. BN-3 RPV-12 11/11/98 Geris GE-UT-700V2 16 Vessel Cire Weld Shell 3 to Shell 4. Geometry and nineteen (19) acceptable flaw indications recorded by Geris GE-UT-70!V2 16 Geris ID system. Coverage was 94.4% -

B-A VCB-BC-5 218" GE. BN-3 RPV-16 11/11/98 Geris GE-UT-700V2 16 Vessel Cire Weld Shell 4 to Flange. Geometry and twenty seven (27) acceptable flaw indication recorded Geris GE-UT-70!V2 16 by Geris ID system. Manual exams NRL Composite coverage was 933% OL UT-CNS-300V3 16 45S UT-CNS-300V3 16 60S UT-CNS-300V3 16 B-A VLA-BA 1 N/A GE. BN-3 RPV-01 10/30S 8 OL UT-CNS-300V3 16 Vessel long Seam Weld Shell 1. No access for Geris 2000 ID system. Manual exams limited due to 45S UT-CNS-300V3 16 nonles, insulation support rings and bioshield access. Coverage is 59.5%. No Recordable Indications 60S UT CNS-300V3 16 (NRI)

B-A VLA-BA 2 N/A GE. BN-3 RPV 02 10/30/98 Geris GE-UT-700V2 16 Vessel Long Seam Weld Shell 1. Geometry and four (4) acceptable flaw indications recorded by Geris GE-UT-70lV2 16 Geris ID system. Manual exams were NRL Composite coverage was 80.6% OL UT-CNS-300V3 16 45S UT-CNS-300V3 16 60S UT-CNS-300V3 16 B-A VI.A-BA-3 N/A GE. BN-3 RPV-03 11/10/98 Geris GE-UT-700V2 16 Vessel Long Scam Weld Shell 1. Geometric and Thirteen (13) flaw indications recorded by Geris ID Geris GE-UT-70lV2 16 system. Two flaw combinations exceeded IWB-3500 and accepted per IWA 3600. Manual exams NRI. OL UT-CNS-300V3 16 Composite coverage was 80.6%. Reference NEDC98-048 45S UT-CNS-300V3 16 60S UT-CNS-300V3 16

e ATTACHMENT 1.1 GE NUCUW?ENERar COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS 1s-Nov.9s Page 2 of 23 ASME COMPONENTID SIZE ISO REPORT DATE FlAM PROCEDURE REV CAL BLK

'B-A VLB-BA-1 N/A GE. BN-3 RPV-05 11/10/98 Geris GE-UT-700V2 16 s, -

Vert:cl Long Scam Weld Shell 2. Seven (7) acceptable flaw indications recorded by Geris ID sptem. Geris GE-UT-701V2 16 Coverage was 80.8%

B-A VLB-BA-2 N/A GE. BN-3 RPV-06 11/11/98 N/A N/A 16 Vessel Long Seem Weld Shell 2. No access for Geris 2000 or manual exams.

B-A VLB-BA-3 N/A GE. BN-3 RPV-07 11/10/98 Geris GE-UT-700V2 16 Vessel long Scam Weki Shell 2. Geometry and twenty two (22) acceptable flaw indications recorded by Geris GE-UT-701V2 16 Geris ID system. Coverage was 49.6%.

BA VLC-BB-1 N/A GE. BN-3 RPV-09 11/11/98 N/A N/A 16 Vessel Long Seam Weld Shell 3. No access for Geris 2000 or manue'. exams.

B-A VLC-BB-2 N/A GE. BN-3 RPV-10 11/12/93 Geris GE-UT 700V2 16 Vesse! Long Seam Weld Shell 3. Geometric and forty nine (49) flaw indications recorded by Geris ID Geris GE-UT-70IV2 16 system. Three flaw combinations exceeded IWB-3500 and accepted per 1WA-3600. Manual exams NRL OL UT-CNS-300V3 16 ComposPc coverage was 98.8%. P 4:rence NEDC98-048 45S UT-CNS-300V3 16 60S UT-CNS-300V3 16 s

B-A VLC-BB-3 gRA GE. BN-3 RPV-11 11/06/98 Geris GE-UT 700V2 16 Venel Long Seam Weld Shell 3. Geometry and thirty six (36) acceptable flaw indications recorded by Geris GE-UT-70!V2 15 Gens ID system. Manual exams NRL Composite coverage was 95% OL UT-CNS-300V3 16 45S UT-CNS-300V3 16 60S UT-CNS-300V3 16 B-A VLD-BB-1 N/A GE. BN-3 RPV-13 10/30/98 Geris GE-UT-700V2 16 Vessel Long Scam Weld Shell 4. Geometry and five (5) acceptable flaw indications recorded by Geris ID Geris GE-UT 70lV2 16 system. Manual exams N'll Composite coverge was 99.5% OL UT-CNS-300V3 16 45S UT-CNS-300V3 16 60S UT-CNS-300V3 16 B-A VLD-BB-2 "/A GE. BN-3 RPV 14 10/30/98 Geris GE-UT-700V2 16 Vessel long Scam Weld Shel! 4. Geometry and five (5) acceptable flaw indicatima recorded by Geris ID Geris GE-UT-701V2 16 rystem. Manel exams NRI. Composite coverage was 99.4% OL UT-CNS-100V3 16 45S UT-CNS400V3 16 60S UT-CNS-300V3 16 s

B-A VLD-BB-3 N/A GE. BN-3 10/30/98 Geris GE-UT-700V2 16 l RPV 15 Vessel Long Seara Weld Shell 4. Geometry and nine (9) acceptable flaw indications recorded by Geris ID Geris GE-UT 70lV2 16 system. Coverage was 98.2% 16 j

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l ATTACHMENT 1.1 GE AUCLEARENERGr COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS 1s-Nov 9s l Page 3 of 23 ASME COMPONENT ID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BLK 18 COMPONENTS FOR CATEGORY B-D B-D NVE-BD-NI A - 28" CE.232 231 R-211 10/1668 OL UT-CNS-300V3 RO 16 NI A Noaje Weld Examined from vessel side for 32.2% coverage. Reference RI-21 45S UT-CNS-300V3 RO 16 60S UT-CNS-300V3 RO 16 B-D NVE-BD-N2C 12" CE.232-231 R-199 10/12 S 8 OL UT-CNS-300V3 RO 16 N2C NouJe Weld Examined from vessel side for 40% coverage. Reference RI-21 45S UT-CNS-300V3 RO 16 60S UT-CNS-300V3 RO 16 B-D NVE-BD-N2F 12" CE.232-231 R-212 10/1768 OL UT-CNS-300V3 RO 16 N2F Nonle Weld Examined from vessel side for 40% coverage. Reference RI-21 45S UT-CNS-300V3 RO 16 60S UT-CNS-300V3 RO 16 B-D NVE-BD-N2O 12" CE.232-231 R-202 10/13/98 OL UT-CNS-300V3 RO 16 N2O NouJe Weld Examined from vessel side for 4tW. coverage. Reference RI-21 45S UT-CNS-300V3 RO 16 60S UT-CNS-300V3 RO 16 B-D 12" CE.232-231 R-203 10/13/98 OL UT-CNS-300V3 RO 16 NVE{D-N2H N2H Nonle Weld Examined from vessel side for 40% coverage. Previously recorded acceptable 45S UT-CNS-300V3 RO 16

! amination recorded with no change noted. Reference RI-21 60S UT-CNS-300V3 RO 16 B-D NVE-BD-N3A 24" 73IE611 R-196 10/10.58 OL UT-CNS-300V3 RO 16 N3A Noale Weld. One rejectable indication recorded. Examined from vessel side for 35.5% coverage. 45S UT-CNS-300V3 RO 16 Reference RI-21 60S UT-CNS-300V3 RO 16 60S UT-CNS-300V3 RO 16 B-D NVE-BD-N4B 12" CE.232-231 R 5 10/14/98 OL UT-CNS-300V3 RO 16 N4B NouJe Weld Examined from vessel side for 31.7% coverage. Reference RI-21 45S Uf-CNS-300V3 RO 16 60S UT-CNS-Z *)OV3 RO 16 B-D NVE-BD-N4D 12" CE.232-231 R-208 10/15/98 OL UT-CNS-300V3 RO 16

- N4D Nonle Weld, one previously recorded rejectable indication recorded. Acceptable per IWA-3600. 45S UT-CNS-300V3 RO 16 Examined from vessel side for 31.~% coverage. Reference RI-21 60S UT-CNS-300V3 RO 16 B-D NVE-BD-N7 6" CE.232-244 R-215 10/21/98 OL UT-CNS-300V3 RO 15 N7 NonJe Weld Examined from head side for 58.6% coverage. Reference RI-21 45S UT-CNS-300V3 RO 15 60S UT-CNS-300V3 RO 15 B-D NVIR-BD-NIA 28" CE.232-231 R-210 10/16/98 Z1 UT-CNS-311V4 RO 16 NI A Nonle Inner Radius Exam 100% Coverage Z2A UT-CNS-311V4 RO 16 B.D NVIR-BD-N2C 12" CE.232-231 R-198 10/12/98 Z1 UT-CNS-311V4 RO 16 N2C NonJe Inner Radius Exam 100% Coverage Z2A UT-CNS-311V4 RO 16 l

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1 ATTACHMFNT E l GE NUCLEARENERor COOPER NUCLEARi 4 TION Unit 1 REIS ABSTRACT OF EXAMINATIONS 1s-Nov-9s Page 4 of 23 ASME COMPONENT JD SIZE ISO REPORT DATE EXAM P. <IEDURE REV CAL BLK B-D NVIR-BD-N2F 12" CE.232 231 R-213 10/17 S 8 Z1 UT-CNS-311V4 RO 16

)

N2F Nozzle Inner Radius Exam 100% Coverage Z2A UT-CNS-311V4 RO 16 Il-D NVIR-BD-N2O 12" CE.232 231 R-200 10/13/98 Z1 UT-CNS-311V4 RO 16 N2O Nozzle Inner Radius Exam 100% Coverage Z2A UT-CNS-311V4 RO 16 j

B-D NVIR BD-N211 12" CE.232-231 R-201 10/13/98 Z1 UT-CNS-311V4 RO 16 l

N211 Nozzle Inner Radius Exam 100% Coverage Z2A UT-CNS-311V4 RO 16 B-D NVIR-BD-N3A 24" GE.731E611 R-190 10/08/98 Z1 UT-CNS-311V4 RO 16 N3A Nozzle Inner Radius Exam NRI Z2A UT-CNS-31IV4 RO 16 Z2A UT-CNS-311V4 RO 16 l

B-D NVIR-BD-N4B 12" CE.232-231 R-207 10/14/98 Z1 UT-CNS-311V4 RO 16 l N4B Nozzle Inner Radius Examined for ASME Sec. XI and Nureg 0619100% Coverage achieved Z2B UT-CNS-311V4 RO 16 Z2A UT-CNS-311V4 RO 16 B-D NVIR-BD-N4D 12" CE.232-231 R-209 10/15/98 Z1 UT-CNS-311V4 RO 16 N4D Nozzle Inner Radius Examined for ASME Sec.XI and Nureg 0619100% Coverage achieved Z2B UT-CNS-311V4 RO 16 Z2A UT-CNS-311V4 RO 16 B-D NVIR-BD-N7 6" CE.232-244 R-219 10/21/98 PT PT-CNS-100V1 R1 N/A N7 Nozzle Inner Radius. Liquid Penetrant exam from component ID.

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1 ATTACHMENT 1.1 l GE NUCLEARENEnGr COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS 1s.Nov-9s Page 5 of 23 i ASME COMPONENTID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BLK 2 COMPONENTS FOR CATEGORY B-D EXP l 1

~

B-D EXP NVE-BD-N3B 24" 731 E611 R-218 10/24/98 OL UT-CNS-300V3 RO 16 N3B Noule Examined per IWB-2430 scope expansion, Reference RI-27. 35.5% coverage due to Nozzle 45S UT-CNS-300V3 RO 16 j configuration. Reference RI-21 60S UT-CNS-300V3 RO 16 )

B-D EXP NVE-BD-N3D 24" 731E611 R-217 10/245)8 OL UT-CNS-300V3 RO 16 N3D Nozzle Examined per IWB-2430 scope expansion, Reference RI-27,35.5% coverage due to Nozzle 45S UT-CNS-300V3 RO 16 configuration. Reference RI-21 60S UT-CNS-300V3 RO 16 L

ATTACHMENT 1.1 GE NUCUMRENERar COOPER NUCLEAR STATION Unit 1 I RE18 ABSTRACT OF EXAMINATIONS 1s-Nov-9s Page 6 of 23 ASME COMPONENT 1D SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BLK 5 COMPONENTS FOR CATEGORY B-F B-F CSB-BF 1* 13.52" CNS-CS-3 R 176 10/10/98 PT PT-CNS-100V1 R1 N/A NSB Safe End to Nozzle examinul with Smart 2000 system. Non-relevant indications recorded. Safe 45S UT-CNS-209V2 RO 51 End taper limits coverage to 95%. 45S UT CNS-209V2 RO 61 45L UT-CNS-209V2 RO 121 60L UT-CNS-209V2 RO 121 OL UT-CNS-102VI RO 51 OT UT-CNS-102VI RO 61 B-F RAS-BF-1 29" CNS-RR-37 R-183 10/22/98 PT PT-CNS-100V1 R1 N/A NI A Nozzle to Safe End examined with Smart 2000 system. Non-relevant indications recorded. Safe 45S UT-CNS-209V2 R0 57 End taper limits coverage to 99%. 45S UT-CNS-209V2 R0 59 45L UT-CNS-209V2 R0 121 60L UT-CNS-209V2 R0 121 OL UT-CNS-102Vi R0 57 OT UT-CNS-102VI R0 59 B-F RRC-BF-1 14.44" CNS-RR-38 R 180 10/14/98 PT PT-CNS-100V1 R1 N/A N2C Safe End to Nozzle examined with Smart 2000 system. Non-relevant indications recorded. Safe 455 UT-CNS-209V2 RO 52 End taper limits coverage to 97%. 45S UT-CNS-209V2 R0 60 45L UT-CNS-209V2 R0 121 60L UT CNS-209V2 R0 121 B-F RRE-BF-1 14.44" CNS-RR-38 R-178 10/12/98 PT PT-CNS-100V1 R1 N/A N2E Safe End to Nozzle examined with Smart 2000 system. Non-relevant indications secorded. Safe End 45S UT CNS-209V2 RO 60 tcper limits coverage to 96% 45S UT-CNS-209V2 RO 52 45L UT-CNS-209V2 R0 121 60L UT-CNS-209V2 R0 121 OL UT-CNS-102V1 R0 60 OT UT-CNS-102VI RO 52 B-F RRH-BF-1 14.44" CNS-RR-37 R-181 10/23/98 PT PT-CNS-100V1 R1 N/A N211 Safe End to Nozzle examined with Smart 2000 system. Welding discontinuities and non-relevant 45S UT-CNS-209V2 R0 52/60 indications recorded. Safe End taper limits coverage to 97%. 45L UT-CNS-209V2 R0 121 60L UT-CNS-209V2 R0 121 45S UT-CNS-209V2 R0 52/60 35L UT-CNS-209V2 R0 121 70L UT-CNS-209V2 R0 121 OT UT-CNS-102VI R0 52/60

ATTACHMENT 1.1 GE NUCIE4RENERar COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS 1s-Nov-9s Page 7 of 23 ASME COMPONENT ID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BLK 36 COMPONENTS FOR CATEGORY B-G-1 B-G-1 PRB-BGl-I to 18 6" CE232-239 R-192 10/20S 8 VT-1 VT-CNS-10lV0 R.2 N/A RPV Nuts 1-18. Acceptable damage to 7,9 & 15. Photos in data package. VT-1 VT-CNS-101VO R.2 N/A B-G-1 PRC-BGl-1 to 18 6" CE232-239 R 193 10/20/98 VT-1 VT-CNS-101VO R.2 N/A RPV Washers 1 18. Normal wrar observed VT-1 VT-CNS-101VO R.2 N/A I

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! ATTACHMENT 1.1 GE NUCLEARENERGY COOPER NUCLEAR STATION Unit 1 l RE18 ABSTRACT OF EXAMINATIONS Js-Nov 9s Page 8 of 23 ASME COMPONENT ID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BLK 10 COMPONENTS FOR CATEGORY B-J B-J CSB-BJ-2* 10" 2502-1 R-177 10/10/98 PT PT-CNS-100V1 R1 N/A NSB Pipe to Safe End Weld, examined with Smart 2000 system. Non-relevant indications recorded. 45S UT-CNS-208V2 RO 49 60L UT-CNS-208V2 R0 49 OT UT-CNS-102VI R0 49 B-J CSB-BJ-3* 10" 2502 1 R-191 10/09 S 8 PT PT-CNS-100V1 R1 N/A i Pipe to Pipe Manual UT exam NRI 45S PD1-UT-2 R.B 49 60L PDI-UT-2 R.B 49 OT UT-CNS-102V1 49 B-J FWB-BJ-1

12" 2509-1 R-186 10/17/98 MT MT-CNS-100V1 R2 N/A N4B Pipe to Safe End examined with Smart 2000 system. Non-relevant indications and thennal sleeve 45S UT-CNS-208V2 R0 83 geometry recorded. 60L UT-CNS-?08V2 R0 83 OT UT-CNS-106VI R0 83 B-J FWB-BJ 111- 13.94* 2509-1 R-185 10/24/98 MT MT-CNS-100V1 R2 N/A N4B Safe End to Nozzle examined with Smart 2000 system. Non-relevant indications recorded Manual 45S UT-CNS-209V2 R0 89 pickup due to thermocouples. Composite coverage 100%. Credit for Zone 5 Nureg 0619 exams. 60L UT-CNS-209V2 R0 89 45S UT-CNS-106VI R0 89 l BJ FWD-BJ-l

12" 2509-2 R-188 10/1788 MT MT rNS-100V1 R2 N/A N4D Pipe to Safe End examined with Smart 2000 system. Non-relevant indications recorded. 45S UT-CNS-208V2 R0 83 60L UT-CNS-208V2 R0 83 B-J FWD-BJ-111 13.94" 2509-2 R-187 10/16/98 MT MT-CNS-100V1 R2 N/A N4D Safe End to Nozzle examined with Smart 2000 system. Non-relevant indications recorded. Manual 45S UT-CNS-209V2 R0 89 pickup due to thermocouples. Composite coverage 100%. Credit for Zone 5 Nureg 0619 exams. 60L UT-CNS-209V2 R0 89 45S UT-CNS-106V1 R0 89 B-J MSB-BJ-111 24" 731E611 R 189 10/07S 8 MT MT CNS-100V1 R2 N/A N3B Nozzle to Safe End Manual UT exam. NR1 45S UT-CNS-106VI R0 115 OT UT-CNS-106V1 R0 115 B-J RAS-BJ-2 28" CNS-RR-37 R-184 10/2168 PT PT-CNS-100V1 R1 N/A 45S UT-CNS-208V2 R0 56 I N1 A Safe End to Pipe examined with Smart 2000 system. Non-relevant indications recorded.

60L UT-CNS-208V2 R0 56 OT UT-CNS-102V1 RO 56 B-J RRE-BJ-2 12" CNS-RR-38 R-179 10/13/98 PT PT-CNS-100V1 R1 N/A j I N2E Pipe to Safe End examined with Smart 2000 system. Non-relevant indications recorded. 45S UT-CNS-208V2 R0 50 60L UT-CNS-208V2 R0 50 OT UT-CNS-102VI R0 50 l

e ATTACHMENT 1.1 GE NUCIEARENERGr COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS 1s-Nov.9s Page 9 of 23 ASME COMPONENT SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL ID BLK B-J RRII-BJ-2 12" CNS-RR-37 R-182 10/15/98 PT PT-CNS-100V1 R1 N/A N211 Pipe to Safe End examined with Smart 2000 system. Non-relevant indications recorded. 45S UT-CNS-208V2 R0 50 60L UTCNS-208V2 R0 50 OT UT-CNS-102VI R0 50 i,

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ATTACHMENT 1.1 GE NUCIEARENEnor COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS is-Nov 9s Page 10 of 23 ASME COMPONENTID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BLK 1 COMPONENTS FOR CATEGORY B-K-1 B-K-1 PSA-BKl.19 10" 2506-1 R 194 10/20/98 MT MT-CNS-100V1 R2 N/A Pipe Support removed per WO 97-2464 to provide 100% coverage. NRI l

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ATTACHMENT 1.1 GE AT/CLEARENERoy COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS 1s-rov.9s Page 11 of 23 l

ASME COMPONENT ID SI7I ISO REPORT DATE EXAM PROCEDURE REV CAL l BLK 1 COMPONENTS FOR CATEGORY C-F-2 C-F-2 IIPEX-CF-69 20x6 2614-3 R-204 10/14/98 MT MT-CNS-100V1 R2 N/A 20x6 Branch Connection NRI l

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r ATTACHMENT 1.1 GE NUCLE 4RENERGr COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS 1s-No 9s Page 12 of 23 ASME COMPONENT ID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BlK 13 COMPONENTS FOR CATEGORY E-A-Aug.

E-A-Aug DRYWELL* N/A N/A VT-98003 11/04S8 VT-3 VT-CNS-103VI R.0 N/A Pre-exam of area at *270 degrees designated area nine A. Ref. PIR 2-04153 E-A-Aug DRYWELL' N/A N/A VT-98004 11/04/98 VT-3 VT-CNS-103VI R.0 N/A Post-exam of area at *270 degrees designated area nine A. Ref. PIR 2-04153 E-A Aug DRYWI1L* N/A N/A VT-98005 11/04/98 VT-3 VT CNS-103VI R.0 N/A Pre-exam of area at *270 degrees designated area nine. Ref. FIR 2-04153 E-A-Aug DRYWELL* N/A N/A VT-98006 11/0468 VT-3 VT-CNS-103VI R.0 N/A Pod estam of area at *270 degrees designated area nine. Ref. PIR 2-04153 E-A-Aug DRYWELL* N/A N/A VT-98007 11/04/98 VT-3 VT CNS-103VI R.0 N/A Pre-exam of area at

170 degrees designated area nine. Ref. PIR 2-04154 E-A-Aug DRYWELL* N/A N/A Vf-98008 11/04/98 VT-3 VT-CNS 103V1 R.0 N/A Post-exam of area at

170 degrees designated area nine. Ref. PIR 2-04154 E-A-Aug DRYWELL' N/A N/A VT-98009 11/04/98 VT-3 VT-CNS-103VI R.0 N/A Pre-exam of area at *150 degrees designated area nine. Ref. PIR 2-04154 E-A-Aug DRYWELL' N/A N/A VT-98010 11/04/98 VT-3 VT-CNS-103VI R.0 N/A Post-exam of area at

150 degrees designated area nine. Ref. PIR 2-04154 E-A-Aug DRYWELL' N/A N/A VT-980ll 11/04/98 VT-3 VT-CNS-103V1 R.0 N/A Pre-exam of area at *340 degrees designated area nine 16. Ref. PIR 2-04153 E-A-Aug DRYWELL* N/A N/A VT-98012 11/04/98 VT-3 VT-CNS-103VI R.0 N/A Pre-et:am of area at *340 degrees designated area nine 16. Ref. PIR 2 04153 E-A-Aug l TORUS

N/A N/A VT-97001 1/19/98 VT-3 VT-CNS 103VI R.0 N/A Augmented IWE examination performed on

Torus Drywell penetration between Bent 6 & 7. Exams performed pre and post painting. SAT.

E-A-Aug TORUS

N/A N/A VT-98001 10/17/98 VT-3 VT-CNS-103VI R.0 N/A Post painting exam on

Torus under valve SW-AOV-TCV451 A. Flaking, blistering, pealing and other signs of distress observed. Ref. PIR 3-50922 E-A-Aug TORUS

N/A N/A VT-98002 10/17/98 VT-3 VT-CNS-103VI R.0 N/A Post painting exam on

Torus under valve SW-AOV-TCV451B. Flaking, blistering, pealing and other

- signs ofdistress observed. Ref. PIR 3 50922 L

l ATTACHMENT 1.1 GE NUCIEAREVERGr COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS 1s-Nov.9s Page 13 of 23 l

l ASME COMPONENTID SIZE ISO REPORT DATE EXAM PROCFDURE REV CAL I

ULK 1

l 2 COMPONENTS FOR CATEGORY E-G l

50 BOLT / NUT

N/A N/A R.IWE-1 3/07/98 VT1 VT-CNS-10lV0 R.2 N/A i

PSI exam perfonned on *20 bolts and 24 nuts for the Drywell Equipment Hatch. SAT.

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E-O IWE-BOLTING

N/A N/A R-300 11/10/98 VT-1 VT-CNS-101V0 R.2 N/A PSI exam performed on *2 bolts and 2 nuts for the Drywell Equipment IIstch. SAT.

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ATTACHMENT 1.1 GE NUCLEARENERGr COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS 1s-Nov-9s Page 14 of 23 ASME COMPONENTID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL B{

2 COMPONENTS FOR CATEGORY GL-PT GL-PT RWCU-13 4' 2605-3 R-214 10/1988 45S PDI-UT-2 R.B 110 Pipe to Elbow NRI G.L.8841 exam 70L PDI-UT-2 R.B 110 GL-PT RWCU-26 4" 2605-1 R-195 10/19/98 45S PDI-UT-2 R.B 110 Pipe to Elbow Root geometry recorded. G.L. 88-01 exam. 70L PDI-UT-2 R.B 110 OT UT-CNS-102VI R.O 110 I

p ATTACHMENT 1.1 GE NUCLEARENERGY COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS mv-9s Pa t of23 ASME COMPONENTID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL I BLK 10 COMPONENTS FOR CATEGORY GL88-01 GL88-01 CSB-BF-l

13.52" CNS-CS-3 R-176 10/10/98 PT PT-CNS-100V1 R1 N/A N5B Safe End to Nozzle examined with Smart 2000 system. Non-relevant indications recorded. Safe 45S UT-CNS-209V2 R0 51 End taper limits coverage to 95%. 45S UT-CNS-209V2 RO 61 45L UT-CNS-209V2 R0 121 60L UT-CNS-209V2 R0 121 l OL UT-CNS-102VI R0 51 l OT UT-CNS-102VI R0 61 i

GL88-01 CSB-BJ-2* 10" 2502-1 R 177 10/10S 8 PT PT-CNS-100VI R1 N/A l I l

NSB Pipe to Safe End Weld, examined with Smart 2000 system. Non-relevant indications recordei 45S ITF-CNS-208V2 R0 49 !

l 60L UT-CNS-208V2 RO 49 i

OT UT-CNS-102VI R0 49 GL88-01 CSB-BJ-3

10" 2502 1 R-191 10/09 S 8 PT PT-CNS-100V1 R1 N/A )

l Pipe to Pipe NRI 45S PDI-UT-2 R.B 49 60L PDI-UT-2 R.B 49 OT UT-CNS-102VI 49 GL88-01 RAS-BF-1 29" CNS-RR-37 R-183 10/22/98 PT PT-CNS-100V1 R1 N/A NI A Nozzle to Safe End examined with Smart 2000 system. Non-relevant indications recorded. Safe 45S UT-CNS-209V2 R0 57 End taper limits coverage to 99%. 45S UT-CNS-209V2 R0 59 45L UT-CNS-209V2 R0 121 60L UT-CNS-209V2 R0 11.1 OL UT-CNS-102VI R0 57 OT UT-CNS-102VI R0 59 GL88-01 RAS-BJ-2 28" CNS-RR-37 R-184 10/21/98 PT PT-CNS-100V1 R1 N/A NI A Safe End to Pipe examined with Smart 2000 system. Nun-relevant indications recorded. 455 UT-CNS-208V2 R0 56

~

60L UT-CNS-208V2 R0 56 OT UT-CNS-102VI R0 56 GL88-01 RRC-BF-1 14.44" CNS-RR-38 R 180 10/14/98 PT PT-CNS-100V1 R1 N/A N2C Safe End to Nozzle examined with Smart 2000 system. Non-relevant indications recorded. Safe 45S UT-CNS-209V2 R0 52 End taper limits coverage to 97%. 45S UT-CNS-209V2 RO 60 45L UT-CNS-209V2 R0 121 60L UT-CNS-209V2 R0 121 GL88-01 RRE-BF 1 14.44" CNS-RR 38 R-178 10/12/98 PT PT-CNS-100V1 R1 N/A

! N2E Safe End to Nozzle examined with Smart 2000 system. Non-relevant indications recorded. Safe End 45S UT-CNS-209V2 RO 60 I taper limits coverage to %%. 45S UT-CNS-209V2 R0 52 45L UT-CNS-209V2 R0 121 60L UT-CNS-209V2 R0 121 OL UT-CNS-102VI RO 60 OT UT-CNS-102VI R0 52

F 1 1

! ATTACHMENT 1.1 aE uucLEARENERGY COOPER NUCLEAR STATION Unit I l i RE18 ABSTRACT OF EXAMINATIONS is.xov.9s !

Page 16 of 23 i

ASME COMPONENT ID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BLK 10 COMPONENTS FOR CATEGORY GL88-01 GL88 01 RRE-BJ-2 12" CNS-RR 38 R 179 10/13/98 PT PT-CNS-100V1 R1 N/A I N2E Pipe to Safe End examined with Smart 2000 system. Non-relevant indications recorded. 45S UT-CNS-208V2 R0 50 60L UT-CNS-208V2 R0 50 OT UTCNS-102V1 R0 50 GL88 01 RRH-BF-1 14.44" CNS-RR-37 R 181 10/23/98 PT PT-CNS-100V1 R1 N/A )

N2fl Safe End to Nozzle examined with Smart 2000 system. Welding discontinuties and non-relevant 45S UT-CNS-209V2 R0 52/60 indications recorded. Safe End taper limitscoverage to 97%. 45L UT-CNS-209V2 R0 121 !

60L UT-CNS 209V2 R0 121 45r UT-CNS-209V2 RO 52/60 l

35S UT-CNS-209V2 R0 121 70L UT-CNS-209V2 R0 121 OT UT-CNS-102VI R0 52/60 GL88-01 RRIl-BF-2 12" CNS-RR-37 R-182 10/15/98 PT PT-CNS-100V1 R1 N/A N211 Pipe to Safe End examined with Smart 2000 system. Non-relevant indications recorded. 45S UT-CNS-208V2 R0 50 60L UT-CNS-208V2 R0 50 OT UT-CNS-102VI RO 50 1

ATTACHMENT 1.1 ce NUcistasurnar COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS 1s-Nov 9s Page 17 of13 ASME COMPONENT ID SIZE ISO REPORT DATE EXAM PROCEiURE REV CAL BLK 10 COMPONENTS FOR CATEGORY NRO619 NRC519 IWB BJ-l I1 13.94* 2509-1 R-185 10/24/98 MT MT-CNS-100V1 R2 N/A N4B Safe End to Nonle examined with Smart 2000 system. Non-relevant indications recorded. Manual 455 UT-CNS-209V2 R0 89 pickup due to thermocouples. Composite coverage 100%. Credit for Zone 5 Nureg 0619 exams. 60L UT-CNS-209V2 R0 89 45S UT-CNS-106VI R0 89 NR0619 FWD-BJ-l 11 13.94" 2509-2 R 187 10/16/98 MT MT CNS-100V1 R2 N/A N4D Safe End to Nov.le examined with Smart 2000 system. Non-reles snt indications recorded. Manual 45S UT CNS-209V2 R0 89 pickup due to thermocouples. Composite coverage 100%. Credit for Zone 5 Nureg 0619 exams. 60L UT-CNS-209V2 R0 89 45S UT CNS-106VI R0 89 NR0619 NB-N4A 12" CE. 232-243 R-206 10/14SS Z3 UT-CNS-311V4 R0 16 N4 A Nonle Bore Examined for Nureg 0619100% Coverage achieved WR0619 NB-N4B 12" CE. 232-243 R-207 10/14/98 Z3 UT-CNS-311V4 R0 16 N4B Nouje Bore Examined for Nureg 0619 Zone 5 exam performed in data report R-185.100% Z5 UT-CNS-209V4 R0 89 Coverage achieved NR0619 NU-N4C 12" CE. 232-243 R-197 10/10S 8 Z3 UT-CNS-311V4 R0 16 .

1 N4C Nonle Bore Examined for Nureg 0619100% Coverage achieved NR0619 NB-N4D 12" CE. 232-243 R-209 10/15/98 Z3 UT-CNS-311V4 R0 16 N4D Nonle Bore Examined for Nureg 0619 Zone 5 exam performed in data report R-187.100% Z5 UT-CNS-209V2 R0 89 Coverage achieved NR0619 NVIR-BD-N4A 12" CE. 232-231 R-206 10/14 S 8 Z1 UT CNS-311V4 R0 16 N4A Nonle Inner Radius Examined for Nureg 0619100% Coverage achieved Z2B UT-CNS-311V4.R0 16 Z2A UT-CNS-311V4 R0 16 NRO619 NVIR BD-N4B 12" CE. 232-231 R 207 10/14 S 8 Z1 UT-CNS-311V4 R0 16 N4B NouJe Inner Radius Examined for ASME Sec.X1 and Nureg 0619100% Coverage achieved Z2B UT-CNS-31IV4 R0 16 Z2A UT-CNS-311V4 R0 16 NR0619 NVIR-BD-N4C 12" CE. 232-231 R 197 10!!0S 8 Z1 UT-CNS-311V4 R0 16 N4C Noule Inner Radius Examined for Nureg 0619100% Coverage achieved 2B UT-CNS-311V4 R0 16 2A UT-CNS-31IV4 R0 16 MR0619 NVIR-BD-N4D 12" CE. 232-231 R-209 10/15 S 8 Z1 UT-CNS-311V4 R0 16 N4D Nonle Inner Radius Examined foi ASME SeeX1 and Nureg 0619100% Coverage achieved Z2B UT-CNS-31IV4 R0 16 Z2A UT-CNS-311V4 R0 16

7 ATTACHMENT 1.1 GE NUCILIRENERor COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS is-Nov-9s Page 18 of 23 i

ASME COMPONENTID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BLK 75 COMPONENTS FOR CATEGORY REC REC 2848-14 W25 16.0" CNS-REC-41 R-173 09/22/98 70S UT-CNS IllVO R.0 62 i

Flinge to Elbow Tack weld geometry at 0 90 180 and 270 Deg. 45S UT-CNS-111V0 R.0 62 l OT UT CNS-106VI R.0 62 {

60RL UT-CNS-111VO R.0 62 l

REC 2848-!4-W35 4.0" CNS-REC-41 R-127 09/14 S 8 70S UT-CNS-111VO R.0 127 P;pe to Pipe ID Weld Ocometry 45S UT-CNS-111VO R.0 127 REC 2848-14-W36 4.0" CNS-REC-41 R 128 09/14 S 8 70S UT-CNS-111 V0 R.0 127 Elbow to Pipe NRI 45S UT-CNS-111VO R.0 127 REC 2848-14-W37 4.0* CNS-REC-41 R 129 09/14/98 70S UT-CNS-111V0 R.0 127 Pipe to Elbow NRI 45S UT-CNS-111VO R.0 127 REC 2848-14-W38 4.0" CNS-REC-41 R-130 09/14 S 8 70S UT-CNS.111V0 R.0 127 Elbow to Pipe NRI 45S UT-CNS-11IVO R.0 127 REC 2848-14-W39 4.0" CNS-REC 41 R 131 09/14 S 8 70S UT-CNS-111V0 R.0 127 Pipe to Elbow NRI. 45S UT-CNS-111VO R.0 127 REC 2848-14 W66 4.0" CNS-REC-41 R 132 09/14SB 70S UT-CNS-111VO R.0 127 Pipe to Pipe NRI 45S UT-CNS-111VO R.0 127 REC 2848-14-WE 4.0" CNS-REC-41 R 133 09/22S 8 70S UT-CNS-Il1VO R.0 127 Valve to Pipe NRI 45S UT-CNS-111VO R.0 127 IGC 2848-15-W12 3.0" CNS-REC-42 R ll3 09/2398 70S UT-CNS-111V0 R.0 122 Reducer to Elbow NRI 45S UT-CNS-111VO R.0 122 i

REC 2848-15-W13 3.0" CNS-REC 42 R il4 09/23S 8 70S UT-CNS-IllVO R.0 122 Elbow to Elbow NRI 45S UT-CNS-111VO R.0 122 REC 2848-15-W14 3.0" CNS-REC-42 R 115 09/23/98 70S UT-CNS-111VO R.0 122 Elbow to Pipe NRI 45S UT-CNS-111VO R.0 122 REC 2848-15-W34 4.0" CNS-REC-42 R-134 09/23/98 70S UT-CNS-111V0 R.0 127 Reducer to Pipe NRI 45S UT-CNS-111VO R.0 127 REC 2848-16-W10 3.0" CNS-REC-44 R-116 09/1768 70S UT-CNS-111V0 R.0 122 Elbow to Pipe NRI 45S UT-CNS-111V0 R.0 122

ATTACHMENT 1.1 GE NUCLEARENERGr COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS is-Nov.ps Page 19 of 23 ASME COMPONENT ID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BLK REC 2848-16 W29 4.0" CNS-REC-43 R 151 09/17/98 70S UT-CNS-IllVO R.0 127 Pipe to Pipe NRI 45S UT-CNS-111VO R.0 127 REC 2848-16-W35 3.0" CNS-REC-44 R-117 09/1768 70S UT-CNS-111VO R.0 122 Elbw to Pipe NRI 45S UT-CNS-111VO R.0 122 )

REC 2848-16-W58 3.0" CNS-REC-43 R-118 09/16 S 8 70S UT-CNS-I l lVO R.0 122 Pipe to Pipe NPJ 455 UT-CNS-111V0 R.0 122 REC 2848-16-W59 2.5" CNS-REC-43 R-101 09/16 S8 70S UT-CNS-IllVO R.0 123 Pipe to Elbow NRI 45S UT-CNS-11IVO R.0 123 REC 2848-16-W60 2.5" CNS-REC-43 R-102 09/16S 8 70S UT-CNS-111V0 R.0 123 Pipe to Elbow NRI 45S UT-CNS-111VO R.0 123 REC 2848-16-W61 2.5" CNS-REC-43 R-103 09/16/98 70S UT-CNS-I l lVO R.0 123 Elbw to Pipe NRI 45S UT-CNS-111VO R.0 123 REC 2848-16-W62 2.5" CNS-REC-43 R-104 09/16 S 8 70S UT-CNS-111VO R.0 123 Pipe to Elbow NRI 45S UT-CNS-11IVO R.0 123 REC 2848-16-W63 2.5" CNS-REC-43 R-105 09/16S 8 70S UT-CNS-111VO R.0 123 Elbw to Elbow NRI 45S UT CNS-111VO R.0 123 REC 2848-16-W64 2.5" CNS-REC-43 R-106 09/1688 70S UT-CNS-111V0 R.0 123 Elbw to Pipe NRI 45S UT-CNS-111VO R.0 123 REC 2848-16-W82 3.0" CNS-REC-43 R 119 09/16 S 8 70S UT-CNS.I11VO R.0 122 Pipe to Pipe NRI 45S UT-CNS-111V0 R.0 122 REC 2848-16-W83 3.0" CNS-REC-43 R-120 09/16S 8 70S UT-CNS-111VO R.0 122 Pipe to Pipe NRI 45S UT-CNS-111V0 R.0 122 REC 2848-16-W9 3.0" CNS-REC-43 R-121 09/1768 70S UT-CNS-111V0 R.0 122 Pipe to Elbow NRI 45S UT-CNS-111VO R.0 122 REC 2848-16-WE 3.0" CNS-REC-43 R-122 09/17/98 70S UT-CNS-111V0 R.0 122 Pipe to Elbow NRI 45S UT-CNS-111VO R.0 122 REC 2848-16-WM 2.5" CNS-REC-44 R-107 09/14 S 8 70S UT-CN -111V0 R.0 123 Pipe to Pipe NRI 87% coverage due to adjacent pipe 45S UT-CNS-111 V0 R.0 123

ATTACHMENT 1.1 GE NUCLEARENERar COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS is-rov.9s Page 20 of 23 ASME COMPONENTID SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL BLK REC 2848-16-WN 2.5" CNS-REC-43 R-108 09/14/98 70S UT-CNS-Il1VO R.0 123 Pipe to Pipe NRI 45S UT-CNS-111V0 R.0 123 REC 2848-16-WO 4.0" CNS-REC-44 R-135 09/15/98 70S UT-CNS-111V0 R.0 127 Pipe to Pipe NRI 45S UT-CNS-111V0 R.0 127 REC 2848-2-W18 12.0" CNS-REC-37 R-165 09/1768 70S UT-CNS-111V0 R.0 130 Tee to Pipe NRI 45S UT-CNS-111V0 R.0 130 i

REC 2848-2-W22 12.0" CNS-REC-37 R-166 09/15 S 8 70S UT-CNS-111VO R.0 130 Tec to Pipe NRI 45S UT-CNS-111VO R.0 130 -

REC 2848-2-W22A 12.0" CNS-REC-37 R 167 09/15/98 70S UT-CNS-l l1VO R.0 130 Pipe to Pipe NRI 45S UT-CNS-11IVO R.0 130 REC 2848-2-W23 12.0" CNS-REC-37 R-168 09/15 S 8 70S UT-CNS-IllVO R.0 130 Pipe to Elbow NRI 45S UT-CNS-111VO R.0 130 ,

l REC 2848-2-W24 12.0" CNS-REC-37 R-169 09/15/98 70S UT-CNS-111VO R.0 130 Elbw to PipeID geometry 45S UT-CNS-11IVO R.0 130 OT UT-CNS-106V1 R.0 130 REC 2848-2 W25 12.0" CNS-REC-37 R-170 09/17S8 70S UT-CNS-IllV0 R.0 130 Pipe to Elbow NRI 45S UT-CNS-111V0 R.0 130 REC 2848-2-W26 12.0" CNS-REC-37 R 171 10/30/98 70S UT-CNS-Il1VO R.0 130 Elbw to Pipe ID geometry. 45S UT-CNS-111VO R.0 130 OT UT-CNS-111VO R.0 130 REC 2848-2-W27 12.0" CNS-REC-37 R-172 09/15/98 70S UT-CNS-111V0 R.0 130 Pipe to Flange NRI 45S UT-CNS-111VO R.0 130 OT UT-CNS-106V1 R.0 130 REC 2848-2 W63 4.0" CNS-REC-37 R 136 09/23 S 8 70S UT-CNS-IllVO R.0 127 Pipe to Elbow NRI 45S UT-CNS-111V0 R.0 127 ;

REC 2848-2-W64 4.0" CNS-REC-37 R 137 09/23 S 8 70S UT-CNS-111V0 R.0 127 Elbw to Pipe NRI 45S UT-CNS-111VO R.0 127 REC 2848-2-W65 4.0" CNS-REC-37 R-138 09G3/98 70S UT-CNS-111 V0 R.0 127 Pipe to Pipe NRI 45S UT-CNS-111VO R.0 127

ATTACHMENT 1.1 GE NUCIEARENERGY COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS is-Nov 9s Page 21 of 23 I

ASME COMPONENTID SI72 ISO REPORT DATE EXAM PROCEDURE REV CAL !

BLK REC 2848-2-W7 6.0" CNS-REC-37 R 156 09/10S 8 70S UT-CNS-111V0 R.0 126 Fbnge to Reducer Backing Ring Geometry 45S UT CNS-111VO R.0 126 OT UT-CNS-106VI R.0 126 REC 2848-2-W73 4.0" CNS-REC-37 R-139 09/15 S 8 70S UT-CNS-111VO R.0 127 Pipe to Reducer NRI 45S UT CNS-111VO R.0 127 i

REC 2848-2-W74 6.0" CNS-REC-37 R-157 09/15/98 70S UT-CNS-Il1V0 R.0 126 !

l Reducer to Pipe NRI 45S UT CNS-111VO R.0 126 REC 2848-2-W74A 6.0" CNS-REC-37 R-174 09/15S 8 70S UT-CNS-111VO R.0 126 Branch Connection Weld NRI 45S UT-CNS-111V0 R.0 126 REC 2848-2-W76A 12.0" CNS-REC-37 R-175 09/15S 8 70S UT-CNS-111VO R.0 126 Branch Connection Weld NRI 45S UT-CNS-111V0 R.0 126 REC 2848-2-W77 6.0" CNS-REC-37 R-158 09/15S 8 70S UT-CNS-111V0 R.0 126 Elbw to Pipe NRI 45S UT-CNS-111VO R.0 126 REC 2848-2-W77A 6.0" CNS-REC-37 R-159 09/1568 70S UT-CNS-111V0 R.0 126 Pipe to Pipe NRI 45S UT-CNS-111V0 R.0 126 REC 284B-2 W9 8.0" CNS-REC-37 R-161 09/10S 3 70S UT-CNS-111VO R.0 128 Pipe to Elbow NRI. 85% coverage due to welded stanchion. 45S UT-CNS-111V0 R.0 128 REC 2848-2-W99 10.0" CNS-REC-37 R-164 09/15S 8 70S UT-CNS-I llVO R.0 129 Flange to Elbow NRI 45S UT-CNS-111VO R.0 129 REC 2848-2-WAA 6.0" CNS-REC-37 R-160 09/15!98 70S ITf-CNS-111VO R.0 126 Vulve to Elbow NRI 45S UT-CNS-111V0 R.0 126 REC 2848-2-WG 8.0" CNS-REC-37 R-162 09/10S 8 70S UT-CNS-Il1V0 R 0 128 l

Elbw to Valve NRI 45S UT CNS-111VO R.0 128 i i

REC 2848 2-WII 8.0" CNS-REC-37 R 163 09/10S 8 70S UT-CNS-111V0 R.0 73 Valve to Valve NRI 45S UT-CNS-111VO R.0 73 REC 2848-2-WZ 6.0" CNS-REC-37 R-155 09/15 S 8 70S UT-CNS-111V0 R.0 126 Pipe to Valve NRI 45S UT-CNS-111VO R.0 126 REC 2848-21-W4 2.5" CNS REC-45 R-109 1 0/30/98 70S UT-CNS-111VO R.0 123 l

ElbwtoPipe NRI 45S UT-CNS-111VO R.0 123

ATT ACHMENT 1.1 GE NUCLEAREVERGY COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS is-Nos ps Page 22 of 23 ASME COMPONENT SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL ID BLK REC 2848-21-W8 2.5" CNS-REC-45 R-110 9/14S8 70S UT-CNS-Il1VO R.0 123 Reducer to Pipe NRI 45S UT-CNS-111VO R.0 123 REC 2848-21-W9 2.5" CNS-REC-45 R-111 09/14 S 8 70S UT CNS-111VO R.0 123 l I

Ellmw to Pipe NRL Welded attachment 100% coverage 45S UT-CNS-111 VO R.0 123 l l

2.5" REC 2848-21-WAC CNS-REC-45 R-112 09/14 S 8 705 UT-CNS-I llVO R.0 123 Pipe to Elbow NRI 92% Coverage due to inner radius of cibow 45S UT-CNS-111VO R.0 123 REC 2848-50-WJ 4.0" CNS-REC-40 R-143 09/23/98 70S UT-CNS-111 VO R.0 127 Pipe to Pipe NRI 45S UT-CNS-111V0 R.0 127 REC 2848-56-W10 3.0" CNS-REC-50 R 123 09/09 S 8 70S UT-CNS-l llV0 R.0 122 ;

I Finnge to Pipe ID Root Geometry 45S UT-CNS-111VO R.0 122 OT UT-CNS-106VI R.0 122 REC 2848-56-W31 3.0" CNS-REC-50 R-124 09/09/98 70S UT-CNS-111VO R.0 122

' Flange to Pipe NRI One side exam. 455 UT-CNS-111 VO R.0 122 I REC 2848-56-W8 3.0" CNS-REC-50 R-125 09/09/98 70S UT-CNS-111V0 R.0 122 Elbow to Flange NRI One side exam. 45S UT-CNS-111VO R.0 122 REC 2848-56-W9 3.0" CNS-REC-50 R-126 09/09/98 70S UT-CNS-111V0 R.0 122 Pipe to Elbow Backing Ring Geometry 45S UT-CNS-111V0 R.0 122 OT UT-CNS 106V1 R.0 122 REC 2848-9 W18 4.0" CNS-REC-40 R 140 09/2168 70S UT-CNS-111VO R O 127 Pipe to Elbow NRI 45S UT-CNS-111VO R.0 127 REC fo48-9-W20 4.0" CNS-REC-40 R-141 09/21/98 70S UT-CNS-111VO R.0 127 Pipe. ElbowNRI 45S UT-CNS-111V0 R.0 127 REC 2848-9 W21 4.0" CNS-REC-40 R-142 09/2168 70S UT-CNS-Il1VO R.0 127 ElbowtoPipe NRI 45S UT-CNS-111VO R.0 127 REC 2848-9-W42 4.0" CNS-REC-40 R-144 09/23 S 8 70S UT-CNS-111VO R.0 127 Pipe to Pipe NRI 45S UT-CNS-111VO R.0 127 REC 2848-9 W43 4.0" CNS-REC-40 R-145 09/21S 8 70S UT-CNS-111V0 R.0 127 Pipe to Elbow NRI 45S UT-CNS-111VO R.0 127 REC 2848-9-W5 4.0" CNS-REC-40 R 146 09/21S 8 70S UT-CNS-IllVO R.0 127 Elbow to Pipe NRI 45S UT-CNS-111VO R.0 127

i.

ATTACHMENT 1.1

! as xuctsansusnar COOPER NUCLEAR STATION Unit 1 RE18 ABSTRACT OF EXAMINATIONS is xov.9s Page 23 of 23 l ASME COMPONENT SIZE ISO REPORT DATE EXAM PROCEDURE REV CAL ID BLK f REC 2848 9.W56 4.0" CNS-REC-40 R-147 09/21S 8 70S UT CNS-111VO R.0 127 i

Elbow to Pipe 1 Linear Indication. See PIR 2-20667. Repair exam performed 10/10/98 on Data Sheet C- 45S UT-CNS-111 VO R.0 127 l 101. No evidence ofindication observed. OT UT-CNS-106VO R.0 127 60RL UT-CNS-111 V0 R.0 127 70S UT-CNS-111VO R.0 127 45S UT-CNS-111VD R.0 127 REC 2848-9-W57 4.0" CNS-REC-40 R-148 09/2168 70S UT-CNS-111VO R.0 127 Pipe to Elbow NRI 45S UT-CNS-111VO R.0 127 REC 2848-9-W58 4.0" CNS-REC-40 R-149 09/21/98 70S UT CNS-111VO R.0 127 ElbowtoPipe NRI 45S UT-CNS-111VO R.0 127 REC 2848-9.W86 4.0" CNS-REC-40 R-150 09/21/98 70S UT-CNS-111VO R.0 127 Pipe to Pipe NRI 45S UT-CNS-111VO R.0 127 REC 2848-9-WA 4.0" CNS-REC-40 R-152 09/23/98 70S UT-CNS-111VO R.0 127 Pipe to Pipe NRI 45S UT-CNS-111VO R.0 127 REC 2848-9-WB 4.0" CNS REC-40 R-153 09/2168 70S UT-CNS-111VO R.0 127 PipetoPipe NR1 45S UT-CNS-111VO R.0 127 REC 2848-9-WW 4.0" CNS-REC-40 R-154 09/2168 70S UT-CNS-111VO R.0 127 ElbowtoPipe NRI 45S UT-CNS-111VO R.0 127 I

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ATTACHMENT 2

\

EXAMINATION PERSONNEL :

i i

aw ..m. ., . .m .. . .., &

Attachment 2.1

)

General Electric Examination Pers:nnel f:r RF018 Name Certifications and Level l Bernardy, James M. Levet Il-L, UT Cert. Umited to Thic s Only Bragg, Olen J. Jr. Level: 111 PT, MT, UT

[ Cameron, Richard E. Level: lit, PT, VT-1-2-3, RT f Chr' stensen, Brian J. Level: ll-L UT (Limited tkns. only) Level: l-T PT, MT Clay, Sean P. Level: 11 PT, MT, UT, VT-1 Conti, Adam A. Level. ill PT, MT, UT

! Cribbe, Walter O. Level. Il PT, MT, UT l

Dingman, Daniel A. Level:ll UT Levet I-T PT, MT Dion James G. Levet il UT Levet i PT Levet 1-T MT Gilliard, John C. Levet ul UT l

Hall, Douglas W. Levet ll PT, MT, VT-1 Levct i UT Hancock, David R. Levetil UT Levet 1 PT, MT l Holland, Lowell T. Levet 11 PT, MT, Levet Il-L UT(Limited tkns. only)

Kloe'.er, Harlan J. Levet 11 PT, UT

! Koenig, Dawn Renee Levet 11 PT, MT Levet Il-L UT (Limited tkns only)

Minor, Chris A. Levet lil PT, MT, UT Money, Richard C. Levet ll PT, MT UT, VT-1-2-3 Pemberton, Michael R. Levet 11 PT, Mt, UT, VT-1-2-3 Romano, Joseph V. Levet ll UT Ryder, Jeffery M. Levet il If-L (Umited tkns. Only) Levet i PT, MT Schlortt, Hermann W. Levet lli PT, MT, UT, VT-1-2-3 Spivey, Randall G. Levet il PT, MT, UT, VT-12-3 Page 1 of 1

ATTACHMEN T 2.2 Wistinghzuse Eximination Perzenn:1 for RF018 l Name Certifications and Level ;

f Anderson, Michale J. LevelII - VT-1,3 l

Elhoff, William K. LevelII - VT-1,3 Hobbie, Kyle D.* LevelII - UT Jones, Barry C. Level 11 - VT 1,3 Moreau, Darrell A.

Level 111 - UT l Moreau, Andre W.

LevelII - UT Nagata, Mitchell. Levelli - VT-1,3 Reaves, Larry C. Level 111 -VT-1,3 l:

Rowland, Denzil R. Levelli - VT-1,3 Story, Jerry Level 111 VT-1,3 Zhang, Yingda

Level 111 - UT

Individual performed BWRVIP-18 augmented examinations, and did not perform ASME XI examinations.

Page 1 of 1

I Attachtaant 2.3 NPPD Examination Perstnnd I: for RF018 Name Certifications and Level i Ackerman, Terry L LevelII - VT-1,2,3 l

l Alexander, Glenn R. LevelII - VT-1,2,3

)

Allen, Van E. Level ll - VT-1,2,3 l Anderson, Steven W. LevelII - VT-1,2,3

Bantz, Clinton S. Level 11 - VT-1,2,3 l

l Baruth, weston L - LevelII - VT-2

l. l Bedgood, Robert A. Levelli - VT-2 l

. j Beger, Nathan L. Levelli - VT-2 Behr, Gnarles A. LevelII - VT-2 l

f Behrends, Kirby L. Level 11 - VT-1,2,3 Billesbach, Douglas S. Level ll VT-1,2 Billings, Michael W. Level ll - VT- 2 Bird David M. Level ll - VT- 2 i

Boden, Marlin D Levelli - VT-1,2,3 Borgen, Terry A. Level li - VT- 2 l

Brandt, Donnie E. LevelII - VT.2 l Brattsovsky, Jeffery M Levelli - VT-1,2,3 l

l Bromen, David J. Levelli - VT- 2 I

Carisan, Randel W. LevelII - VT.2 l Carpenter, Marvin W. LevelII - VT-2

, 7 j Carpenter, Allen R. Level 11 - VT- 1 i l Collins, John W. LevelII - VT-2 l

l Conner, Robert L. Leveill - VT 2 Cunningham, Dale A. Level II - VT- 2 Dierberger, Scott W. Level II - VT- 2 l

Domino, Jeffrey F. LevelII - VT-2 Flock, Paul J. Level!! - VT- 1,2,3 Garner, Blair F.- Levelli - VT- 2 Gonnella, Mark E. LevelII - VT-2 Grossman, Tony L. LevelII - VT- 1,2,3 Page 1 of 4 !

t I

LL j

i Attachment 2.3 NPPD Examination Personnel for RF018 l Name Certifications and Level l Hakenwerth, Douglas G. LevelII- VT-2 Halkens, MeMn E. Level II- VT-2 Hall, Allan L. Level 11 - VT- 1,2,3 Hall, James C. LevelII - VT- 1,2,3 Hannaford Jr., Martin D. LevelII - VT-2 l Hansen, Mark S. LevelII - VT-2 Harpham, Joseph C. Leveill - VT-2 Hartman, Vincent J. Level 11 - VT.2 Hasselbring, Brian J. LevelII - VT-2 )

l Hawkins, Hylan, A. LevelII - VT-2 '

[ l Helms, David P. LevelII - VT 2 Herold, Michael K. LevelII - VT- 1,2,3 l

Hitzel, Harry D. LevelII - VT-2 Hoff, Steven D. Level 11 - VT-2

{ Holm, Carl R. Levei ll - VT- 2 Holmes, Mark A. Level II - VT- 2 Hoskins, Randy A. LevelII - VT-2 Jennings ,Kurt,% Level 11 - VT-2 Kahanca, Joh'n R. Level 11 - VT-2 Kaul, Mark S. Levelli - VT-2 King, Keith R. Leveill - VT-2 Kleckinger, Allen D. LevelII - VT- 1,2 l Knopik, Mark B. LevelII - VT- 1,2,3 Kubes, Ricky L Levelli - VT-2 Lavigne, Paul G. Level ll - VT- 2 l Long, Jerry J. Level 11 - VT-2 Maine, Richard L. LevelII - VT-2 Mason, Rory L Leveill - VT.2 McCargill, Donald L. LevelII - VT-2 ,

i McElfresh, Charles E. Level 11 - VT-2 l Page 2 of 4 i

Attachment 2.3 NPPD Examination Personnel for RF010 i

Name Certifications and Level i McKay, Richard L. LevelII - VT-1,2,3 Moody, Mark A. Level ll - VT- 1, 2, 3 Mueller, Timoth, R. Level 11 - VT-2 Murphy, Brian P. Level 11 - VT-2 Nichols, Jesse A. LevelII - VT-2 Norris, Steven P. Level 11 - VT-2 Nosbich, Kenneth L LevelII - VT-2 Pebley, Rodney J. LevelII - VT- 1,2 Penfield, Rod L. LevelII - VT- 2 Perry, Gary L. LevelII - VT-1,2,3 Peters, Marvin E. Levelli - VT- 1,2,3 Peterson, Kristopher S. Leveill - VT.2 l Pope, James W. Leveill- VT-2 Pugh, Timothy J. LevelII - VT- 1,2,3 Rasmussen, James A. LevelII - VT-2 Ratzlaff, Te' .' L. Level 11 - VT-2 Reed Jr., Robert F. Levelli - VT-2 Reeves, Clinton A. Leveill- VT- 1,2,3 l Reimers, Arlie L. LevelII - VT- 1,2,3 Riley, Richard K Leveill - VT-2 Sailors, Robert E. Level 11 - VT- 1,2,3 Saul, Noel S. Levelli - VT-2 l Scheppman, Darrell W. Level ll - VT- 2 Schizas, Fred A. LevelII - VT- 2,3 Schwindt, Warren F. Level 11 - VT-2 Shandy, Wayne E. Level 11 - VT- 1

! Shaw, Ronald L. Level ll - VT-1,2,3 Sherman, Robert L Leveill - VT- 1,2 Siske, William J. LevelII - VT- 1,2,3 Slama, Robert A. LevelII - VT- 2 Page 3 of 4

i; Attachment 2.3 NPPD Examination Personnel for RF018 l Name Certifications and Level 4 l Sienker, Timothy S. Levelli - VT-2 Smalifoot, Steven C. Levelli - VT-2 Smdh, Joel L. Level ll1- VT-1,2,3 Sullwold, Kim D. LevelII - VT- 1,2,3

. Tackett, Michael L Level 11 - VT- 2 l Talmon, Larry L. Leveill- VT-3 l Tanderup, Richard J. Levelli - VT-2 l Tanner, Rick L. Level 11 - VT-1,2,3 Tanner, Kurt LevelII - VT- 1,2,3 l Tetric! PaulM. Level 11 - VT- 2 Thomas, Kenneth B. LevelII - VT-1,2,3 Towne, Wade A. LevelII - VT-2 Tune, David H. LevelII - VT-2 Umshler, Daniel L Level 11 - VT-2 l Vanrenen, Jeffrey L. LevelII - VT- 1 Volk, Robert C. LevelII - VT- 1,2,3 Volker, Justin R. Level 11 - VT. 2 Willaims, Richard K. Level 11 - VT.1,2 Wrtzak, Robert A. Levelli - VT-2

)

Zabel, Charles F.- LevelII - VT- 1,2,3 1

Page 4 of 4

ATTACHMENT 3 EXAMINATION PROCEDURES i

l l

I

ATTACHMENT 3.1 Prretdure Li:t Fcr Ex minitirn3 Performed by General Electric for RFO18 Procedure Revision Title ADM-CNS-1002V2 0 Procedure for Nondestructive Examination Data Review and Analysis of Recorded Indications Procedure for Performing Linearity Checks on Ultrasonic GE-ADM-1001 Version: O Instruments Procedure for Zero Reference and Data Recording for GE-ADM-1005 Version: 0 Nondestructive Examinations Procedure for Training and Qualification of Personnel for GE-NE GE-ADM-1025 Version: 2 Specialized NDE Applications GE-ADM-2032 Procedure for the installation and Removal of the Geris 2000 ID Version: O Manipulator in The RPV Procedure for Assembly, Installation, and Removal of the GERIS GE-ADM-2006 Version: O 2000 ID Lower Guide Ring in The RPV Procedure for The Assembly, installation, and Removal of The GE-ADM-2007 0 GERIS 2000 Upper Guide Ring in The RPV MT-CNS-100V1 Procedure for Magnetic Particle Examination Using the Yoke 2

FRR 1G08L-002 Technique (Dry Particle Color Contrast or Wet Particle, Fluorescent)

Procedure for Liquid Penetrant Examination (Visible Dye, Color PT-CNS-100V1 1 Contrast, or Fluorescent)

Procedure for Manual Ultrasonic Examination of Similar and UT-CNS-102V1 0 ,

Dissimilar Piping Welds UT-CNS-104VO 2 Procedure for Manual Ultrasonic Planar Flaw Sizing Procedure for Manual Ultrasonic Examination of Dissimilar Metal UT-CNS-105V1 0 Nozzle to Safe End Welds Procedure for Manual Ultrasonic Examination of Ferritic Piping and N 40W1 0 Vessel Welds 2" and Less in Thickness Precedure for Manual Ultrasonic Examination of Ferntic Piping UT-CNS-111VO

0 Welds for IGSCC ,

UT-CNS-208V2 0 Procedure for Automated Ultrasonic Examination of Similar and )

Dissimilar Piping Welds Procedure for Automated Ultrasonic Examination of Dissimilar Metal NS-20W2 0 Nozzle to Safe End Welds UT-CNS-300V3 Procedure for Manual Ultrasonic Examination of Reactor Vessel 0

FRR 1GLO8L-001 Assembly Welds UT-CNS-308V1 0 Procedure for Manual Examination of the RPV Threads in Flange Procedure for Manual Ultrasonic Planar Flaw Sizing of Nozzle inner UT-CNS-309V3 0 Radius and Bore Regions

Procedure for Manual Ultrasonic Examination of Nozzle inner UT-CNS-311V4 0 Radius and Bore Procedure for Ultrasonic Demonstration Showing Acoustically UT-CNS-603VO 1 Similar Material Characteristics GE-UT-700 Procedure for The Examination of Reactor Pressure Vessel Welds erskn: 2 FRR 1B8P3-001 Using The GERis 2000 ID Procedure for Reactor Pressure Vessel Flaw Sizing with The GERIS GE-UT-701 Version: 2 2000ID 3 Procehre for VT-1 Examination R G 8L-003 VT-CNS-103V1 0 Procedure for VT-3 Examination PDI Generic Procedure for the Ultrasonic Examination of Austenitic PDI-UT-2 B Pipe Welds PDi Generic Procedure for Ultrasonic Through Walt Flaw Sizing in PDI-UT-3 B Pipe Welds

Procedure used for performance of augmented examinations, and not used for performance of ASME XI examinations.

Page 1 of 1

7 ATTACHMENT 3.2 Proc 2 dure Liit Fcr Ex minitigni Performed by Westinghouse for RFO18 l'

Procedure Number ' Revision Title BWRP-2.2 - 0 with Field Reactor Pressure Vessel Intemals Visual Examination for Cooper Nuclear Station Change No. 5 GCPR-ISI-10

0, with Field Procedure for Performing Ultrasonic Equipment Linearity Verification Change No. 3 l CSP-ISl-100

Procedure for Ultrasonic Examination of BWR Core Spray Assembly Welds Using 3

the WesDyne LT40 or UDRPS System l STD-OP-1996-7684

1 Procedure for Installation and Removal of the Core Spray Piping Scanner l

i

Procedure used for performance of BWRVIP-18 augmented examinations, and not used for performance of ASME XI examinations.

I i

l l

1 I

l l j i

l i

l Page 1 of 1

ATTACHMENT 3.3 I Pr:cidure Li;t Fsr Extmin tinn3 J Performed by NPPD for RF018 Procedure Number Revision Title 3.28.1.1 0 Visual Inspection of Pressure Retaining Bolting and Integral Attachments, VT-1 j l

3.28.1.3 1 Visual Inspection of Pump Casings And Valve Bodies, VT-3 l

3.28.1.4 1 General VisualInspection of Containment Surfaces 3.28.1.5 0 Visual Inspection of Containment Surfaces, VT-1/3 6.HPCI.501 1 ASME Section XI Periodic Pressure Test of the Class 2 High Pressure Coolant injection i

6.NBl.501 2 ASME Section XI Periodic Pressure Test of the

6. MISC.502 4 ASME Class 1 System Leakage Test 6.1RHR.501 3 ASME Section XI Periodic Pressure Test of the Class 2 Residual Heat Removal System 6.RCIC.501 1 ASME Section XI Periodic Pressure Test of the Class 2 Reactor Co e Isolation Cooling 7.0.8.1 8 Inservice Leak Testing 7.0.8 13/14 Pressure Testing I

2.57 7.1 ASME Category F-A Component Supports inspet. ion and Adjustment !

l l

l Page 1 of 1 l

1 ATTACHMENT 4 l

l l

l EQUIPMENT AND MATERIALS l

1 l

l 1

l Attachment 4.1 General Electric Equipment and Materials for RF018 Approved Materials Ba Manufacture . Type

.u r 0928531 Ultrasonic Couplant

Krautkrarner-Brason Exosen 30 9145305 14198301 MT Materials

Magnaflux 8 Red 94a079 Dubi-Chek PT Materials

Sherwin DP-40 721-A1 Penetrant Dubl-Chek Sherwin D-100 731-A6 Developer Dubi-Chek Sherwin DR-60 610-14 Cleaner

Materials supplied by NPPD but used by General Electric.

ApprovM Equipment Manufacture / Model Serial Number Scopes UT Manual Staveley Sonic 136 136P1106C021349 Staveley Sonic 136 136P110SCO31363 Staveley Sonic 136 136P1106C031372 p Staveley Sonic 136 136P1106C031373 Stavelety Sonic 136 689H Staveley Sonic 136 136P-7701 Smart Auto UT Tecrad Tomoscan TTS10092118 Mag Yoke Parker Research PAR-ACMT-026 Lift Block Parker Research CAL-10MT-008 l

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Page 1 of 3

p-Attacht.6

! General Electric Equl> ment and Materiale

! fir RF018 l

l -

Approv uipment ManufactureI Model Serial Number l ..

Thermometers PTC 1200CL PTC 1203CL PTC 1218CL r

Approv uipment identification Number Material Reference Blocks CAL-RHON-003 CS CAL-RHON-008 CS CAL-RHOM-054 CS ..

O CAL-RHOM-011 SS CAL-RHOM-034 CS CAL-DPTH-54 i SS CAL-WILLY-0; i SS l

CAL-ilW2-026 CS i

l-Page 2 of 3

Attachment 4.1 General Electric Equipment and Materia 4 far RF018 Apw utpment Manufacture / Model Serial Number Size Frequency (MHz)

Transducers KBA 04091 .16" x .394" 4.0 KBA 003YMC .25" 2.25 l

KBA B03417 .25" 2.25 KBA B21400 .25" 2.25 KBA D19643 .25" 2.25 K3A LO5759 .25" 2.25 KBA 003YMJ .375" 2.25 KBA 004067 .375" 2.25 KBA 003T8~. .50" 2.25 KBA PA243 .50" 1.5 KBA 0065PB .50 x 1.00" 1.0 KBA B35427 .50 x 1.00" 2.25 KBA E02105 .50" x 1.00" 2.25 KBA K02712 .50" x 1.00" 2.25 VRA L21920 .50" x 1.00" 2.25 KBA C14440 .50" x 1.00" 1.0 KBA J07016 .50" x 1.00" 1.0 KBA 001WWY .75" 2.25 KBA F19493 1.00" 2.25 KBA 0016HD 3.5 x 10 MM 2.0 KBA 00984 10 MM 4.0 Megasonic E0528 2 (14 x 30) 4.0 1

Megasonic E1625 2 (14 x 30) 4.0 RTD 98-143 2 (10 x 16) 1.5 RTD 98-147 2 (10 x 16) 1.5 l RTD 98-151 2 (10 x 16) 2.0 RTD 98-167 2 (10 x 18) 2.0 RTD 98-173 2 (10 x 18) 2.0 RTD 98-187 2 (10 x 18) 2.0 RTD 98-243 2 (10 x 18) 2.0 RTD 98-190 2 ( 6 x 18) 2.0 RTD 94 175 2 (8 x 14) 20 Page 3 of 3

Attachment 4.2 Watinghnuse Equipm;nt End Miterials for RF018 Approvea tquipment manufacturer i Type Model Serial Wh Ultrasonic Scopes Tektronix 2247A

B031035 Wesdyne

Dynapulser 1544 Model 103 Approvea tquipment manufacturer i Type Model Serial Number Size Frequency (MHz)

Transducers Panametrics

119329 .500X000" 2.25 Sigma

2616-98003A 4(8x7)mm 2.25 Sigma

2616-98004B 4(8x7)mm 2.25 l

l Type ID Number Material Cdibration Block WDI-98-001

304 SS WDI-98-002

304 SS WDI-98-003

304 SS 1

Equipment used for performance of BWRVIP-18 augmented examinations, and not used for performance of ASME XI examinations. I l

l Page 1 of 1 L

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l Attachment 4.3 ,

NPPD Equipment and Materials j for RF018 STAMPED PROGRAM IDENTFICATION TAB WAREHOUSE NUMBER IE mlCKNESS MATERIAL NUMBER NUMBER CNS. CAL.STD.NO.15 15 CNS 19186 RPV 4.000" CS CNS. CAL.STD.NO.16 16 CNS 19187 RPV 7.100" CS CNS-49-10-80-SS 49 CNS 19217 0.593" SS CNS-50-12-80-SS 50 CNS 19218 b 0.688" SS CNS-51-13-1.125-SS 51 CNS 19219 1.125" SS CNS-52-14-140-SS 52 CNS 19220 141 1.250" SS j CNS-56-28-1.25-SS 56 CNS 19224 28.0" 1.250" SS CNS-57-29-1.935-SS 57 CNS 19225 29.0" 1.935" SS CNS-59-29-1.620-CS 59 CNS 19227 29.0" 1.620" CS CNS-00-14 .972-CS 60 CNS 19228 14.0" 0.972" CS l I

CNS-61-13 .844-CS 61 CNS 19229 13.0" 0.844" CS CNS-62-16 .375-CS 62 CNS 34546 16.0" 0.375" CS I CNS-73-8-100-CS 73 CNS 20964 8.0" 0.594" CS CNS-83-12-1.00 83 CNS 24954 12.0" 1.000" CS CNS-89-14-100-CS 89 CNS 20974 14.0" 0.938" CS CNS-110-4-80-SS 110 CNS 19233 4.0" 0.337" SS CNS-115-24-1.593 115 CNS 24999 24.0" 1.593" CS CAL-N2NZ-003 121 N/A 12.0" 1.300" CS/INC/SS CNS-122-3 .216-CS 122 CNS 32422 3.0" 0.216" CS CNS-123-2.5 .203-CS 123 CNS 32423 2.5" 0.203" CS CNS-126-6 .280-CS 126 CNS 32637 6.0" 0.280" CS CNS-127-4 .237-CS 127 CNS 33989 4.0" 0.237" CS CNS-128-8 .322-CS 128 CNS 33990 8.0" 0.322" CS CNS-129-10 .365-CS 129 CNS 3991 10.0 0.365 CS CNS-130-12 .375-CS 130 CNS 33992 12.0" 0.406" CS CAL-DEPTH 054 GE N/A N/A N/A SS I

l Page 1 of 1 J

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l ATTACHMENT 5 l

SUBMITTAL OF EXAMINATION RESULTS ACCEPTED BY ANALYTICAL EVALUATION l

l CNS RPV FLAW EVALUATION HANDBOOK CALCULATION No. NEDC98-048 3

l l

H: \ CNSPROCS \ FORMS \ VOL3 \ 3-4-7 \ 3-4-7_1 Nebraska Public Power District DESIGN CALCULATIONS COVER SHEET Calculation No. NEDC98-048 Title CNS RPV Flaw Evaluation Handbook Task identification No. N/A Design Change No. N/A System / Structure NS-VES-RPV Discipline Civil / Structural Classification: I X 1 Essential i 1 Non-Essential l

Calc.

Description:

This calculation incorporates by attachment General Electric's (GE) CNS Reactor Pressure Vessel (RPV) flaw Evaluation Handbook Report Number GENE 523-008-0194, Revision 1, prepared for the inservice inspection (ISI) for key vessel welds per requirements of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. Incorporation made in accordance with Section 8.2 of CNS Procedure 3.4.7. The purpose of GE's CNS Reactor >

Pressure Vessel (RPV) Flaw Evaluation Handbook is to evaluate and to documerrt the results of the ISI Reactor Pressure Vessel key welds to determine if additional evaluation is needed for the welds. The inspection results will be evaluated using the " Flaw Evaluation Procedure

developed by GE in Report Number GENE-523-008-0194 Rev.1 (see Appendix B of Attachment 3.1).

T ISI inspection activity utilized the GERIS 2000 equipment in inspecting the welds during 1998 Refueling Outage i

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l W

//ff '

0 ' Original Issue GE 10/15/98 Wietse/A N#e N/A U 0 \V 1 i h/n/M M14 '

v <

Rev. Status Prepared Checked or Design Approved No. Revision Description By/Date Reviewed By/Date Verification /Date By/Date j Status Codes

1. As Boilt - 3. For Construction

2. Informatior) only 4. Superseded or Deleted

' j

H:\ CNSPROCS \ FORMS \ VOL3 \ 3-4-7 \ 3-4-7_2 Sheet._1 of_.L_

Nebraska Pub!ic Power District DESIGN CALCULATION CROSS REFERENCE INDEX NEDC 98-048 Rev. No. .D__ Prepared By: General Electric (GE) Checked / Reviewed By Ali Racha !

Date: 10/15 19_9fL Date: 11/01 19 98 r- 1

- Rev. PEND N C GES TO DESIGN INPUTS AFFECTED DOCUMENTS 1 Combustion Engineering Analytical -

None None Report Number CENC 1150 2 GE Report Number GE-NE-523- -

None None 159-1292, Febmary 1993 ;

1 3 GE Report Number MDE-103-0986, None Nonc j May 1987

)

4 ASME Boiler and Pressure Vessel None I None Code,Section XI,1989 Edition 5 ASME Boiler and Pressure Vessel -

None None Code, Section IU,1989 Edition

6 Letter from T.A. Caine to -

None None !

M.Ber.nett:" Review of Current !

Cooper P-T Curves Against Methods of Reg. Guide 1.99, Revision 2, May 1992 j l

7 US Nuclear Regulatory 2 None None

. Commission, Reg. Guide 1992, Revision 2, May 1988 l

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g ht, pAgf. I k O O O J

L H:1 CNSMtOCS n FORMS WOL313 4-713 4-7,,3 N1braska Public Power District DESIGN CALCULATIONS SHEET sneet l ot .3_

NEDC 98 048 Prepared By: General Electric (GE) Checked / Reviewed By: Ali Bacha Rev. No. O Date: 10/15 19 98 Date: 11/01 19 98 e

l PURPOSE-

'Ihis calculation incorporates by attachment General Electric's (GE) CNS Reactor Pressure Vessel (RPV)

Flaw Evaluation Handbook Report Number GENE 523-008-0194, Revision 1, prepared for the Inservice Inspection (ISI) for key vessel welds per requirements of American Society of Mechanical Engineers (ASME)

Boiler and Pressure 5 essel Code. Incorporation made in accordance with Section 8.2 of CNS Procedure 3.4.7. The purpose of GE's CNS Reactor Pressure Vessel (RPV) Flaw Evaluation Handbook is to evaluate and to document the results of the ISI Reactor Pressure Vessel key welds to determine if additional evaluation is needed for the welds. The inspection results will be evaluated using the " Flaw Evaluation Procedure" developed by GE in Report Number GENE-523-008-0194 Rev.1 (see Appendix B of Attachment 3.1).

The ISI inspection activity utilized the GERIS 2000 equipment in inspecting the welds during 1998 Refueling Outage (RFOl8).

EXTENT OF REVIEW-General Electric's (GE) CNS Reactor Pressure Ve=sel Flaw Evaluation Handbook was performed under GE's QA program. Therefore, the NPPD review does not include in-depth checks of mathematical calculations, but rather focuses on the general acceptability of design inputs, methodology, and conclusions l to the best of the reviewer's ability. Any significant comments or concerns identified during the review have been resolved and incorporated by GE.

1 REVIEW SUMM ARY- I 1.0 PURPOSE The purpose of GE's "CNS RPV Flaw Evaluation Handbook"(Attachment 3.1) is to evaluate the results of the laservice Inspection (ISI) of the CNS Reactor Pressure Vessel (RPV) for key vessel welds to determine if additional evalu3 tion per ASME Code is needed for the welds. The welds shown on the CNS Partial RPV " Rollout" DrawinE Figure 1-2 of the Handbook are the subject of the ISI inspection during CNS's 1998 refueling outage.

2.0 .RFFFRENCFR i See page 45 of the attached " Flaw Evaluation Handbook".

3.0 KITACHMENTS 3.1- General Electric's " Cooper Nuclear Station RPV Flaw Evaluation Handbook", September 1998, Report Number GENE.523-008-0194 Revision I and GE's Supplement Letter (GE/CNS98018) dated October 27,1998.

H: % CNSPROCS % FORMS i VOL3 % s-4-7 % 3 4-7_3 Nebr ska Public Pow 2r District DESIGN CALCULATIONS SHEET sheet I ot 3 NEDC 98-048 Prepared By: General Electric (GE) Checked / Reviewed By: Ali Bacha ping c9 Y Rev. No. O Date: 10/15 19 98 Tf/fJ Date: 11/f/ 19 98 3.0 A'ITACHMENTS (continue) 3.2 Structural Integrity's Letters reference MLH-98-045, MLH-98-047, MLH-98-056, MLH-98-063, MLH-98-066, and MLH-98-067 3.3 GE's Comment Resolutions (Letters) 3.4 NPPD Intra-District Memos 3.5 Analysis of UT Indications in the Reactor Pressure Vessel (RPV) Shell Welds 3.6 UT Report No. RPV-03 and Corresponding Flaw Evaluation Worksheets 3.7 UT Report No. RPV-10 and Corresponding Flaw Evaluation Worksheets 4.0 CAI CIII ATION INPITTS See the Flaw Evaluation Handbook (Attachment 3.1) for the applicable design inputs.

5.0 ASSIIMPTIONS )

The design inputs utilized by GE are correct and applicable to CNS.

6.0 METHODOI OGY (see also Attachment 3.1) l 6.1 General i A structural flaw evaluation was performed for CNS in accordance with ASME Code Section XI (1989 Edition) for axial (meridional or longitudinal) and circumferential welds in the vessel cylindrical shell region. He analysis assumes the most limiting loadings for test, normal (Level A), upset (Level B), Emergency (Level C) and Faulted (Level D) conditions. The flaw evaluation provided in the GE's report considers current operation at 16 effective full power years (EFPY),

and includes fatigue crack growth and irradiation embrittlement for one additional cycle (18 months) and end oflife (EOL).

6.2 I nading Canditinm The loading conditions considered in the fracture analysis are hydrotest, boltup (flange region),

and LOFWP (loss of feed water pump). He loadings associated with the hydrotest and LOFWP condition are 1) Membrane pressure stresses,2) Weld residual bending stresses,3) Clad residual stress.

GE has indicated that although the P-T curves used for this analysis allow for reactor operation in areas deemed more limiting than those used in their calculations (e.g. at 120*F and 610 psi, or at 90*F and 312 psi), these pressure / temperature conditions were not considered, because reactor operation is not expected to reach these zones on the P-T curve. The flaw calculations used for this report are based on the expectation that Cooper reactor operation (i.e. power ramp-up and rampdown) will never reach the limiting zones suggested above. The CNS pressure test procedure, j

f H: t CNSPZOCS t FORMS t vots t s.4-7 t s4-7_s Ntbraska Public Powtr District DESIGN CALCULATIONS SHEET sheet 3 et 5 NEDC 98-048 Prepared By: General Electric (GE) Checked / Reviewed By: Ah Bacha Rev No. O Date: 10/15 19 98 Date: 11/01 19 98 l 6.2 Landing Conditions (continue) 6 MISC.502, was reviewed and it is documented that the pressure test procedure follows the pressure temperature curves in the Technical Requirements Manual (TRM) and that after 200 psig inspection, i the vessel is heated to the test temperature before it is pressurized to 1005 psig. Operation in the areas to the left of the curve is not permitted. Therefore, GE's assumption is adequate.

6.3 Analysis Method The analysis methods used in the fracture analysis follow those prescribed in ASME Code Section l

XI IWB-3600 (1989 Edition). Applied stress intensity factors, K 1, were developed as a function ('

of the flaw depth ratio, a/t (surface flaw) or 2a/t (subsurface), and aspect ratio, a/L. 'Ihese were compared to the allowable fracture toughness, K ta, reduced by the Section XI safety factor of (10 for hydrotest, to detemiine allowable flaw sizes. An upper bound on allowable flaw size was established at 1/3 depth of the LAS (Low Alloy Steel) wall thickness to ensure that ASME Code Section III primary stress requirements were met. A lower bound for allowable flaw sizes is established by the minimum inspection standards ofIWB-3500. If the flaw does not satisfy this j standard, continued operation may still be justified if the flaw satisfies the IWB-3600 acceptance criteria. For the later case, there is a reinspection requirement imposed by the Code (see ASME Section IWB-3132). If the flaw does not meet the IWB-3600 acceptance criteria, then more specific analysis of the flaw may show continued operation to be acceptable. If the flaw specific analysis to IWB-3600 criteria were not met, it would be possible to show continued operation to be acceptable on the condition that the pressure test temperature be increased. In an extreme case, either flaw removal or weld repair might be necessary (see ASME Section IWA-4000).

7.0 CALCULATIONS Flaws are evaluated and documented per the guidelines of the " Flaw Evaluation Procedure" and the " Cooper Nt: clear Station Flaw Evaluation Worksheet" of Appendix B of Attachment 3.1.

8.0 CONCLUSION

S All evaluated welds meet the applicable ASME Code allowables as presented in the CNS RPV Flaw Evaluation Handbook. No further evaluation is necessary for any of the inspected RPV welds.

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H:1CNSPROCS iFORMSiVOL3134 7134 7 3 N1brtska Public Powtr District DESIGN CALCULATIONS SHEET Sheet 1of; HEDC 98-048 Prepared By: General Electric (GE) Checked / Reviewed By: Ali Bacha Rev. No. O Date: 10/15 19 98 Date: 11/01 19 98 r

ATTACHMENT 3.1 General Electric's " Cooper Nuclear Station RPV Flaw Evaluation Handbook" September 1998 Report Number GENE 523-008-0194 Revision 1 and GE's Supplement Letter (GE/CNS98018) dated October 27,1998 l

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ooco N EDc %-04S ATTACHMElif 3*I PAGE l 0F 79 TECHNICAL SERVICES BUSINESS GENE-523-008-0194 Rev. I GE Nuclear Energy DRF # 137-0010-7 175 Curtner Avenue, San Jose, CA 95125 Class II September 1998 COOPER NUCLEAR STATION RPV FLAW EVALUATION HANDBOOK l l

September 1998 Prepared for Nebraska Public Power District Prepared by GE Nuclear Energy

(

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GENuclear Energy GENE-523-003-0194 Rev.I DRF H 137-0010-7 .

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ooca M N 96-04S I ATTACHMENT - 3a i f

PAGE - 1 .OF- 76 l

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COOPER NUCLEAR STATION RPV FLAW EVALUATION HANDBOOK (

) '

Prepared by: O.- u.

W. Lai, Engineer Structural, Mechanical, and Materials Engineering Verified by: E2 C a L. Schultz, Engineer Inte Structural, Mechanical, and Materials Engineering Approved by: b l

T.A. Caine, Project Manager Stmetural, Mechanical, and Materials Engineering i

1

q GENuclear Energy GENE-323-008-0194 Rev.I DRF # 137-0010 7 i

ooc o _ klGDcA8-o AB l ATTACHMENT -- 3l IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully l

The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the purchase order between Nebraska I Public Power District (NPPD) and GE, and nothing contained in this document shall be construed as changing the purchase order. The use of this information by anyone other than NPPD, or for any purpose other than that for which it is intended under such purchase order is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

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GENuclear Energy GENE-523-008-0194 Rev.1 i DRF H 137 0010-7 Doc e _ medic.98-046 EXECUTIVE

SUMMARY

ATTACHMENT f.1 PAGE 4 _ c.- 77

- A structural flaw evaluation was performed for Cooper Nuclear Station in accordance with ASME Code Section XI (1989 edition) for axial (meridional or longitudinal) and circumferential welds in the vessel cylindrical shell region. The analysis assumes.the most limiting loadings for test, normal (Level A), upset (Level B),

Emergency (Level C) and faulted (Level D) conditions. The flaw evaluation provided in this report considers ement operation at 16 effective full power years (EFPY), and includes fatigue crack growth and irradiation embrittlement for one additional cycle (18 months) and end of life (EOL). In meneral. Inside surface flaws were found to be limitine for the selected vessel shell welds which were annivzed. Ev=h>=tions were also nerformed for subsurface flaws for the se!ae**d weld reeions. Firure I ida=*ih all the weld renions selected for ===lvsis in the handbook.

Previous analyses have shown that, in general, nydrotest and boltup conditions, which involve the combination oflow operating temperatures and high safety factors, are the most limiting operating conditions for vessel welds. These two conditions bound all load cases for test, normal, upset, emergency, and faulted conditions, with the exception to a Loss of Feedwater Pump (LOFWP) transient, which introduces a rapid drop in temperature for surface welds in the middle to lower vessel region. Hydrotest, boltup (flange regions), and LOFWP conditions were, therefore, the only cases considered for fracture analysis. Based on the P-T curve for 16 EFPY, the minimum specified hydrotest temperature is 195'F at 1100 psig.

Loading associated with the hydrotest or LOFWP condition were:

Membrane pressure stresses

. Weld residual bending stresses Clad residual stress (clad thickness = 5/16 in. nominal) on the inside surface only.

The analysis methods follow those prescribed in ASME Code Section XI IWB-I 3600. Applied stress intensity factors, K I, were developed as a function of the flaw depth 3

GENuctrar Energy _

GENE-523-008-0194 Rev.I DRF # 137-0010-7 \

ooc , NETv %-843 i ATTACHMENT EI PAGE 6 0F M l ratio, a/t (surface flaw) or 2a/t (subsurface), and aspect re.tio, a/L. These were compared !

to the allowable stress intensity factor, K la, reduced by the Section XI safety factor of VIO f

for hydrotest, to determine allowable flaw sizes.

An upper bound on allowable flaw size was established at 1/3 depth of the LAS wall thickness to ensure that ASME Code Section III primary stress requirements were

{

met. A lower bound for allowable flaw sizes is established by the minimum inspection standards of IWB-3500. If the flaw does not satisfy this standard, continued operation may still be justified if the flaw satisfies the IWB-3600 acceptance criteria, as developed in this report. However, for the latter case, there is a reinspection requirement imposed by the ASME Code.

The allowable flaw curves presented in this report are based on conservative assumptions of possible loadings. If a specific flaw were to be found, a more detailed analysis may show larger allowable flaw sizes. A relative ranking of weld regions is shown in Table I. These rankings are based upon the minimum allowable flaw depth at each weld for an aspect ratio of a/L=l/6. The corresponding weld locations are identified in Figure L l 4

GENuclear Energy GENE-323 008-0194Rev.!

DRF M 137-0010-7 DOC # bW%-6&

ATTACHMENT - 3, I TABLEI "3 - - OF 76 Relative Ranking of Weld Regions Based on Allowable Flaw Sizesa for Aspect Ratio of a/L=l/6 = 0.17 Surface Flaws Ranking Weld Region Flaw Depth [in] Flaw Type (See Figure I) (including clad) 1 V2b o,450c axial 2 VI b o,450c axial 3 H12 b o,450c axial 4 VFWinside 0.450c circumferential 5 VFW outside 0.450c circumferential 6 V3 0.865 axial 7 V4 1.073 axial 8 H23 1.316 axial 9 H34 1.590 axial aAllowable flaw sizes can be increased by increasing the hydrostatic test temperatures in most cases.

bBeltline welds cIWB-3500 evaluation limiting - see Section 3.1.1.

/

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l GENuclear Energ GENE-523-008-0194 Rev.1 DRF M 137-0010-7 000 a _N EDc.V-o46 ATTACHMENT - 3, i PAGE- ~7 __ or __ 76 f Top Head Head flange 6 :

VfW Vessel flange V4 H34 Vessel Shell V3 H23 l -

CORE V2 9eltline Region H12 1

VI b - , . ))

y .-

Bottom Head l Note: Not to Secle l FIGURE I. Cooper Nuclear Station weld regions for flavt evaluation.

6

I-l' GENuclear Energy t

GENE-523-008-0194 Rev.1 i

DRF # 137-0010-7 coc e _NEDe.9S o46 i

Table of Contents ArrAcHuEur Li t

PAGE- 4 - OT - "'M EXECUTIVE

SUMMARY

....... ........ _ ... ... .. .. . ... 3

1.0 INTRODUCTION

. ... . ._. . . ._ ...... . . . .. .. . ._... . . ._ .. .. ._....._....8 2.0 ANALYSIS METHODS ..... ... . . .. .. .. . . .. ...... .. .. ... .. .. ..... . .-... .. ._- ! 1 2.1 Assumed Loading . .. .. ... ... .. . . ... . ... ._ ... .. ......_ .... ... ... .. ....... l 1 2.1.I Cladding Residual Stresses. . ...... . .-.. ....... .... .......... .11 2.1.2 Pressure Stresses ...... . . .. . . . ..

._... .... .. ......... ... .... .. 12

2. I 3 Weld Residual Bending Stress.. .... .- .. . ... . .. .. . ..._ .., . ... _ ._.. ..... 13 2.2 Section III Local Membrane Stress Limits . ._..-. ....._ ... . . ...... ... . . . ..13 23 Section XI Fracture Margin Assessment. . . .. . . . -- ..._............I4 23.1 Fracture Toughness and Allowable Stmss Intensity Factors .-.. ... ... .. .. .. 14 2.3.2 Meth, 's Specific to Irradiated (Beltline) Region- ... ......- .14 23J Applied Stress Intensity Factors . ._....... .. . ....... .. .. . . ._ .. 6 '

2.3.4 Allowable Flaw Depths (not including fatigue)...... . ... .. .. . ....... . .. ..18 23.5 Fatigue Crack Growth Allowances. .. .. .. ..........__ ... ..... . ...... ._ . ._. _19 3.0 FLAW ACCEPTANCE DIAGRAMS .. .. ....._... .. ... . .. . . . ..... ..... ... .... . . .... 25 3.1 Flaw Accap8=M Criteria.. ... .... .. . ... . ......_ .. . . ................25 3.1.1 IWB-3500 Acceptance Standards for Examination.. ... .. ......~...............25 3.1.2 IWB-3600 Analytical Acceptance Criteria....... . ... _... . . . . . . . . _ . . . .26 3.2 Development of Acceptance Diagrams ... . . . .... . ..._ ... ..........__.....26 3.3 Relative Ranking of Weld Regions . . . ... . ... . ._................28 4.0

SUMMARY

_ . ........... ...... .... ... . . ...._.. . ... ... .. .. _.. . _......__.......-.43

5.0 REFERENCES

.. .. ... . ...... .. ... ........ . _ .. _ .._ . ._..-. . .. 45 APPENDIX A - Matenals Data Book ar.d Calculations .. .. _ . . _ .. . - -.. . ..~....... .... .. .... .. 47 A 1. Vessel Geometry... ... ... .. _ ... .. ....... . . ~ . . ......._.............48 A2. Limiting InitialRTndt --- - - - - - ~ ~ - - - - - - - -- + - 48 A3. Limiting RTndt for Non-Beidine Regions . . _ .. . . _ . .... . _.. . ... ... ....... 48 A4. Adjusted Reference Temperatures (ART) for Beltline Regions _.. .. . . .... ...... . .. ..... 48 A4.1 Chemistry Data _ ...... .. .... . _ . . . . . . . . _ . . . . . . . . . ..... . 4 8 A4.2 Fluence Information_... . . . . . . . . . . . . . . . . .._...............49 A4J Calculation ofARTndt - - ~ ~ - - - -- - -- ~ ~~ ~~- -- - * ~~.50 A4.4 Margin... . . .. .. _.... -- _ . . . . . . . . - . . . . . . . . . 5i A4.5 Calculation of ARTS .. . . _ . . - ...........................5i AS. Fracture Toughness, Kg, _._._.__. ....~_ ....__ ..._ __._.___.._..__~52 A6. Vessel Test Pressures and Metal Temperatures ..... ... . . . . . . . . ...... 52 A7. Bolt Preload Stresses . . ._ . . . . . . ......... ......._ . . . .52 A8. Cladding Residual Stresses . .. . ...... . . . . . . . . . . . . . _ . . . . . . 52 APPENDIX B - Flaw Evaluation Procedure... - . . . . . . . . . .. .. 59 Bl. Determine Region and Orientation of Flaw .. .. . . ~ . ..... ... .. 60 B2. Flaw Geometry and Classification .. . ... . _ . _ . . . . . ... .. 61 B3. Flaw Size and Aspect Ratio._.. . . . . . . . . . . . . 62 B4. IWB-3500 Flaw Evaluation._... .... ...... ... . ....... _. .. 62 B5. IWB.3600 Flaw Evaluation. . . . . . . . . . . . .. . . _ _ . . 62 B6.Section III Evaluation (1/3 Limit).. .... _.. . . ... .

.. .. . .. 62 B7. Further Evaluation.. . . . . . . . . . . . . . ........... . . .., . 63 7

GENuclear Energy GENE-523 008-0194 Rev.1 DRF # 137-0010-7 000 * $ZU-dM

1.0 INTRODUCTION

cH This report documents a generic flaw evaluation to determine allowable flaw sizes for key vessel welds at Cooper Nuclear Station. The disposition time of any flaw indications found during inspections can be significantly minimized when the fracture mechanics assessment has been performed in advance.

The scope of this report includes the following:

Assurned loading conditions, including pressure stresses, boltup stresses, weld residual stresses and clad residual stresses.

=

Analysis, per IWB-3500 and IWB-3600 Section XI of the ASME Code

[Ref.1J, of allowable surface and subsurface flaw sizes in the various weld regions, taking into consideration conservative estimates of fatigue crack growth and irradiation embrittlement for one additional operating cycle j (18 months) and end oflife (EOL). Current operation is taken to be at 16 EFPY.

Flaw acceptance diagrams showing allowable surface and subsurface fisw depths (a or 2a, respectively) versus aspect ratio (a/L) for all selected weld i

regions and flaw orientations.

Procedure, including a flowchart and worksheet, for evaluating potential flaws found during inspections.

Relative ranking of all selected welds in '.erms of allowable flaw sizes based on an aspect ratio of a/L=l/6.

A flaw evaluation was performed for vessel welds in the beltline region as shown in Figure 1-1. Figure 1-2 shows a " roll-out" of the Cooper vessel with the location and identification of welds shown. The beltline region includes welds VI, V2 and H12.

Although only a portion of weld VI extends into the beltline region, the entire weld is conservatively classified a:: a beltline weld. The labeling convention established in Figure 1-1 and 1-2 is consistently used throughout the handbook to identify specific welds l

or groups of welds.

8

GENuclear Energy GENE-323-008-0194 Rev.)

DRF # 137-0010-7 Doc , EEIE90-o41 ATTACHMENT - 3. I f Top Head PAGE_lO .OF 76 l

i' Head flange b

VfW Vessel fionge V4 H34 Vessel Shell v3 H23 l -

CORE y2 Beltline Region H12 I

VI l

I O

\\ ( ) //

Bottom Head l Note: Not to Scale l FIGURE 1-1. Cooper Nuclear Station vessel weld regions selected for fisw evaluation.

9 g

I GENuclear Energy GENE-523-008-0194 Rev.1 DRF N 137-0010-7 00c , UEDc%-o4B ATTACHMENT 3I P4 GE - II 0F '79 Azimuthal Location 0

30 6,0 90 1,20 150 1,80 240 9 2,10 2,70 3,00 33,0 g g . . . . . . . . . . .

g VFW Weld V4a Weld V4b Weld

[2-234A [2-234B [V4c2-234C Weld 1134 Weld l

[V3ai 234A Weld ["V3bI 234B Weld [V3e 1-234C Weld 1123 Weld e

' e e

a

[V2a1233A Weld Beltline Region

[V2b 1233B Weld [V2c 1233C Weld e

E :

.. ........[........

Via Weld

. . . . . .H.12.W.e l.d . . . . . . . . . . . . . . . . .

a Vib Weld

[2-233A [2-233B [Vic2-233C Weld FIGURE 1-2 Cooper Nuclear Station Partial RPV " Rollout" Drswing

, 10

GENuclear Energy GENE-323-008-0194 Rev.)

DRF # 137-0010-7 coc e AElrAG--o46 2.0 ANALYSIS METHODS AM^ cum - Li PAGE - I2- 0F- ~7 5" 2.1 Assumed Loading Stresses in the region of each weld are assumed to be due to (i) clad residual stress, (ii) pressure stress, (iii) weld residual stress, and (iv) boltup stresses (which were considered at the VFW region). The applied stresses are summarized in Table A-5 of Appendix A.

For the selected vessel welds, previous analyses have shown that boltup and hydrotest conditions are typically the most limiting for fracture mechanics assessment.

Because the system hydrotest and boltup conditions involve a combination of high safety factor and low metal service temperature, resulting fract te toughness values, K la, are significantly lower than almost all other conditions. The exception is Cooper's Loss of Feedwater Pump transient, which results in a rapid temperature drop in the middle to lower vessel region.

For welds adjacent to the closure flange region, preliminary analysis has shown the limiting load condition to be the hydrotest condition, for which the higher test temperature is more than offset by the higher membrane stresses caused by additional pressure loading.

For the remainine vessel welds. nrevious analysis for a similar vessel IRef,21 has shown that bolt nreload stresses are fully attenuated at locations away from the flanne reeion. Therefore, the limiting load condition is either the hydrotest condition or the LOFWP transient. A hydrotest pressure of 1100 psig at 195'F, as determined from the P-T curves for a hydrotest at 16 EFPY [Ref. 3], is used for the analysis.

2.1.1 Cladding Residual Stresses After a stainless steel (SS) clad is applied to the low alloy steel (LAS) vessel shell plate, a post-weld heat treatment (PWHT) is performed at approximately 1150 F to relieve residual stresses. Consequently, cooling below the PWHT temperature results in residual clad stresses because of the difference in the coefficients of thermal expansion between the SS clad and LAS material. Cooling after PWHT causes tensile stresses in the clad which can reach yield level (30-40 ksi) at room temperature. Rybicki, et.al., [Ref. 4] have shown that the clad residual stress at room temperature, following PWHT and shop hydrotest, is 11

]

[ -- _1 000 # ME 98-d4A GENuclear Energy #""*"W

?

,1--

GENE-323-008-0194 Rev.!

DRF H 137 0610-7 approximately equal to the clad yield strength in both hoop and axial directions. Rybicki, et.al., [Ref. 4] used a clad yieid strength equal to 32 ksi at 70 F. Upon subsequent heating,

! the clad stress was found to decrease as a result of thermal expansion. Similar results have j

l been obtained by Ganta, Ayres, and Hijeck [Ref. 5], who have also reported clad residual l stresses on the order of 30 ksi at room temperature, which, because of high temperature l creep effects, are actually lower than the assumed yield strength of 45 ksi. Therefore, based on these results, it would be reasonable and conservative to assume a clad stress '

equal to an assumed yield strength of 35 ksi at 70'F for this aalysis.

To validate this assumption, an analysis (see Appendix A) was performed to show I that, upon cooling from 1150 F to 70*F (room temperature), the elastic residual clad stress does in fact exceed the clad yield strength. Therefore, a clad stress equal to an assumed yield strength of 35 ksi (at 70 F) will be used in this analysis. When the reactor is subsequently heated to the pressure test temperatures, the differences in the thermal expansion coefficients will reduce the tensile stresses in the clad (see Appendix A). At a hydrotest temperature of 195 F, the residual clad stress reduces to 24 ksi due to thermal expansion. This result is consistent with Rybicki, et.al., [Ref. 4], who reported a clad stress of approximately 26 ksi for a 0.125 inch thick clad at 200'F.

Also, to maintain equilibrium, a slight compressive residual stress is induced in the LAS base metal. However, because of the differences in the thickness between the clad and the base metal, the compressive stress in the LAS material is small and will be conservatively neglected.

4 The clad stress is used to compute Kclad to be used in the Section XI fracture mechanics flaw assessment. The clad thickness is nominally 5/16=0.3125 inch at the inside surface of the vessel cylindrical region.

2.1.2 Pressure Stresses Pressurization of the vessel results only in membrane stress. For axial flaws, the hoop stress is calculated according to the thin-walled pressure vessel formulation, PR/t, where R is the inside vessel radius and P is the hydrotest pressure of 1100 psig. This l

12

Doc 0 N b GENuclear Energy A11ACHMLNT 4 u, 5',, GENE-523-008-0194 Rev i ggy y ,37_gg,g_7 l l

l approach is valid for Cooper Nuclear Station, since the limiting R/t =20 >10 in the vessel region. The axial stress for circumferential flaws is calculated using PR/2t.

i 2.1.3 Weld Residual Bending Stress i

Weld residual stress due to the seam weld or the flange weld are reduced significently as a result of PWHT. However, some weld residual stress still remains after PWHT. Based upon previous analysis for seam welds [Refs. 6 & 7], a residual bending I stress of 8 ksi is conservatively assumed in both axial and circumferential directions for flaws oriented parallel to the weld line. This bending stress simulates the measured cosine stress distribution for welds with PWHT [Ref. 6]. For flaws oriented perpendicular to the weld line, the weld residual stress is zero. The LOFWP transient analysis for surface welds i in the middle to lower vessel region, requires an additional weld residal stress of 9 ksi is included [Refs.13 & 14]. This stress is a result of the rapid drop in temperature at the !

inside surface of the vessel wall due to a LOFWP transient. !

2.2 Section III Local Membrane Stress Limits In addition to the fracture mechanics requirements of Section XI, structural !

requirements for primary local stress per Section III [ Ret.' 8] must also be satisfied. The maximum primary local stress cannot exceed 1.5Sm. Since it is assumed that the clad does '

not carry any part of the load, the part of the crack extending into the LAS must be limited to 1/3 the LAS wall thickness. However, for the purposes of this analysis, the net 1/3 wall thickness limit will be conservatively defined as t1/3 limit = 1/3 tLAS (2-l' regardless of flaw classification (i.e. surface or subsurface). For inside surface flaws the 1/3 limit is conservatively measured from the surface of the clad and not fro.a the clad /LAS interface. This limit is used in conjunction with Kallow to determine allowable crack depths as a function of aspect ratio.

13

I l-l 4 GENuclear Energy , , ~ , , . .u GENE-523-008-0194 Rev.)

00co w we ,o-u m 997 y j37,gg39 7 ATTACHMENT Ll PAGE 89 0F M 23 Section XI Fracture Margin Assessment 1

The assumed loads and clad residual stress from Section 2.1 were used to calculate stress intensity factors (KI ) versus crack depth ratio (a/t or 2a/t). Postulated subsurface flaws were conservatively analyzed for the most limiting proximity factors such that the

) allowable flaw depths are bounding for all subsurface flaws. The assumed flaw geometry for surface and subsurface defects are shown in Figure 2-1.

23.1 Fracture Toughness and Allowable Stress Intensity Factors Fracture toughness values, lK a, were calculated per Appendix G of ASME Section XI. K al values are determined for each weld location based upon limiting RTndt values or adjusted reference temperatures (ART), metal service temperatures, and irradiation embrittlement effects (refer to Appendix A). For beltline and non-beltline regions, analyses were performed at the minimum hydrotest temperature of 195"F corresponding to 16 EFPY.

Allowable stress intensity factor limits, Kallow, were determined from the fracture toughness values per IWB-3612 by applying the following safety margin for hydrotest I conditions, j

Kallow= Kla/V10 (2-2)

This value is used to determine the allowable crack depth ratios for each aspect ratio. Note that hydrotest conditions are limiting at all weld locations.

2.3.2 Methods Specifie to Irradiated (Beltline) Region l

Due to irradiation embrittlement, the allowable stress intensity factor will decrease l with increasing EFPYs. This effect is characterized by a shift in RTndt values based upon fluence levels at different EFPYs. The adjusted reference temperatures (ART) are used in place of the initial RTndt values in computing the allowable stress intensity factor as described in Section 2.3.1. The ART values were determined as a function of crack depth I using methods described in US NRC Regulatory Guide 1.9), Revision 2 [Ref.9], (see Appendix A for detailed analysis).

14

GE Nuclear Energy GENE-323-008-0194 hev.1 DRF # 137-0010-7 DOC

N O bb- N ATTACHMENT Me i PAGE IU OF 76 Only the allowable stress intensity factor, Kallow, and fracture toughness are directly affected by irradiation embrittlement. Applied stress intensities (as calculated per Section 2.3.3 below) are not dependent upon irradiation effects, and are only dependent upon applied loading.

15

E L

GENuclear Energy GENE-523-008 9194 Rev.)

DRF # 137-0010-7 i

Doc , MEDc @ -dM i I'* l AT TACHMENT' l 2.3.3 Applied Stress Intensity Factors l

In determining applied stress intensity factors, the following assumptions were made:

l Vessel flaws can be modeled by flat plate analysis as described in Section XI of the ASME Code.

f

. Linear elastic fracture mechanics (LEFM) can be used to determine K I.

The effect of clad stress can be modeled as a point load applied to an edge (infinite) length flaw, which may be subsequently adjusted for finite length i flaws.

A residual bending stress of 8 ksi is conservatively assumed in both axial and circumferential directions for flaws oriented parallel'to the weld line. This bending stress simulates the measured cosine stress distribution for welds with !

PWHT. For flaws oriented perpendicular to the weld line, the weld residual stress is zero.

Although there are several methoas that can be used to determine KI , the LEFM approach for flat plates was used per Appendix A of Section XI from the ASME Code

[Ref.1]. Because of back well bending, the use of flat plate theory has been shown to give conservative results for cylindrical pressure vessels. Since all stresses, with the possible exception of clad stresses, are elastic, LEFM is expected to yield accurate results.

The applied stress intensities for the given stresses are calculated for surface flaw aspect ratios (a/L) and subsurface flaw aspect ratios (2a/L) of 0.0, 0.1, 0.2, 0.3, 0.4, and 0.5 according to the methods given in Appendix A of the ASME Code,Section XI [Ref.1].

L i 16 l-I~

s ,

GENuclear Enery -

GENE-523-008-0194 Rev.I DRF M 137-0010-7 000 s MN98-04e ATTACHMENT 3*I PAGE @ W 'N The stress intensities due to applied membrane and bending stresses are calculated i

per IWB-3600 [Ref.1] as follows:

! Km "Um Mm V(na/Q) (2-3) i Kb "ob MbV(na/Q) (2-4) l where, om = total applied membrane stress ob = total applied bending stress I

a = flaw depth Q = flaw shape parameter-Mm = membrane stress correction factor Mb = bending stress correction factor Although methods for calculating stress intensities for membrane and bending stmsses are included in Section XI, no method is identified for determining Kelad-The clad residual stress is a localized stress that only exists in the clad /LAS interface region. As such, the clad residual stress was modeled as a resultant force (equivalent point load) applied at the mid-thickness of the clad on each face of the crack opening. This model, which is valid only for surface defects, is described by the following equation given by Paris and Sih [Ref.10]:

Km, clad = 2 o clad t elad /V2nc (2-5) where, Km, clad = Kclad as determined for an infinite-length flaw a clad = clad residual stress l

c = a - 0.5 telad-l 17

GENuclear Energy GENE-323-008-0194 Rev.1 DRF # 137-0010-7 000 # odd ATTACHMENT bl This value is then corrected for a finite length crack, Kclad = Km, cladYQm /Q (2-6) where, Q. = shape factor for infinite length flaw Q = shape factor for analyzed finite length flaw The shape factors are identical to those obtained from Section XI of the ASME Code for membrane and bending stresses. The net effect of clad stresses for subsurface flaws is insignificant.

The individual stress intensity contribtitions due to membrane stress, bending stress, and clad point load stress can be combined by method of superposition, as shown in Figure 2-2. The applied stress intensity factor is, therefore, calculated as the sum of individual stress intensity factors, I Kg = Km + Kb+ Kelad (2-7)

This calculation is performed for each aspect ratio as a function of flaw depth (i.e. a/t or 2a/t) to determine the limiting flaw depths (riot including fatigue allowances).

2.3.4 Allowable Flaw Depths (not including fatigue)

To illustrate how allowable flaw depths are calculated, a sample result showing KI versus crack depth ratio is shown in Figure 2-3. The limiting flaw depth ratio (a/t) is defined at the point where the applied stress intensity is equal to the allowable stress intensity, provided that the upper bound 1/3 wall thickness limit is not exceeded. Similar example results are shown in Figure 2-4 for subsurface flaws.

18

GENuclear Energy GENE-523-008-0194 Rev.1 DRF H 137-0010-7 DOC # N EI X:S $ -04 b ATTACHMENT 3+l PAGE M OF I 2.3.5 Fatigue Crack Growth Allowances The limiting crack depths calculated above do not include fatigue crack growth, and must therefore be adjusted to include an additional allowance for crack growth. Cooper is assumed to be currently at 16 EFPY. Consideration of fatigue crack growth is given for one additional operating cycle (18 months) and end oflife (EOL). 1 A fatigue crack growth allowance is conservatively calculated for the limiting flaw I sizes. Since the limiting flaw sizes will have an applied stress intensities equal to or less than the allowable stress intensity, the applied stress intensity range, AK, used to compute fatigue crack growth will be conservatively based upon a maximum stress intensity level of Kmax=Kallow:

AK = Kmax - Kmin (2-8)

Kallow - 0 The minimum stress intensity is conservatively assumed to be Kmin-0. Fatigue crack i growth for all inside surface flaws is calculated for a reactor water environment. The limiting fatigue crack growth rate (for a stress intensity ratio of R = Kmin/Kmax>0.65) is conservatively computed as follows [Ref.1],

i da/dN = 0.252(AK)l.95(p-in/ cycle) (2-9) where AK = maximum stress intensity factor range (ksiVin).

Fatigue crack growth for outside surface flaws and subsurface flaws is computed based upon an air environment [Ref.1]:

da/dN = 0.0267(10-3)K3.726 (p.in/ cycle) (2-10) 19

GENuclear Enerv GENE-323-008-0194 Rev.I DRF # 157-0010-7 DOC # MA@ -dM 3*I ATTACHMEN' PAGE 7 L 0F 6 Approximately 10 cycles per year is conservatively assumed. Currently, Cooper Nuclear Station is assumed to be at 16 EFPY, Therefore, for an additional open ting cycle (18 months), fatigue crack growth allowances were calculated relative to 1.5 l'.FPY. In general, for 15 fatigue cycles at the maximum stress intensity ranges, fatigue crack growth was found to be small - not exceeding 0.01 inch of crack growth. For EOL, fatigre crack growth was calculated relative to 16 additional EFPY.

It was found that the differences between fatigue crack growth values calculated for one operating cycle and those calculated for EOL were small. For this reason, allowable flaw size charts shown in Figures 3-1 through 3-13 consider only the allowable flaw sizes calculated using EOL fatigue crack growth values.

l The fatigue crack growth allowances, Aafcg or 2Aareg, are used to adjust the limiting flaw sizes obtained from the LEFM analysis to develop the acceptance criteria.

I

, 20

r GENuclear EurO GENE-323-008-0194 Rev.!

DRF H 137 0010 7 000

M h 9 8 M 8 ATTACHMENT LI PAGE 72 0F '7I N

-2a - 1/3 t y F t - LAS #

a

-.-- L - T u F

Subsurface Flaw d h t = LAS + Clad LAS Y

-a F

E p ------

h n n 1/3 t - r jad --- '

-= L z Surface Flaw FIGURE 2-1. Flaw geometry for typical planar surface and subsurface defects. l i

z j

2 21 j i

d

GENuclear Energy GENE-523-003-0194 Rev.1 DRF # 137-0010-7 000 # O E'D C @ ~d b ATTACHMENT Mel

x' PAGE 23 - Of 7f

=

= .

==*

t j K e

_=

_ .A.

+

x o

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had

=

~

f

.w x

FIGURE 2-2. Method of superposition for fracture mechs' tics problem.

Suess intensities due to localized clad stresses, membrane stress, and bending stress can be combined using this method.

l 22 l

GENuclear Energy agyg.323.ggg,g,94 p,, ,

DRF M 137 0010 7 Doca h NI-O O ATTACHMENT 3* l PAGE M OF 78 70 . . . . .

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s i.

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a.E a.2 0.4 e.s Flaw Deput Rado. a/t.

Figure 2-3 Sample Results of Kg vs a/t for a Surface Flaw i

l l

I 23 J

GENuclear Energy GENE-323-008-0194 Rev.I DRF # 137-0010-7 DOC a __k)E.DC98-o%

ATTACHMENT &t FAGE _ 2 k or_ yf 78 g

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0.8 a.1 a.2 a.s 84 u Maw Depth Ratio. 2a/L f

Figure 2-4 Sample Results for Kg vs 2a/t for Subsurface Flaws i

l l

24

m ..

l GENuclear Energy GENE-523-008-Ol94 Rev.I DRF M 137-0010-7 000

h 98-04B 3.0 FLAW ACCEPTANCE DIAGRAMS ^*C N - 3

PAGE 2-(F OF- ~78

- 3.1 Flaw Acceptance Criteria 3.1.1 IWB-3500 Acceptance Standards for Examination The ASME Code,Section XI, IWB-3510.1 [Ref.1], outlines the standards for examination for surface and subsurface planar flaws in pressure retaining vessel weld regions. It should be noted that the IWB-3500 acceptance standards for surface flaws are measured relative to the LAS/ clad interface. This is because any flaw found ehtirely within the clad is considered acceptable per IWB-3500. Since the inspection standards specified in the Code do not include clad thickness, the allowables for inside surface flaws are adjusted such that the allowable flaw depth may be measured relative to the surface of the clad rather than the LAS/ clad interface.

Subsurface allowables were calculated based upon a proximity factor of Y = S/a = 1.0, which is typical of a mid-plane flaw. For flaws where Y < 0.4, the flaw must be classified as a surface defect acconting to the proximity rules ofIWA-3300. The j IWB-3500 curves in Figures 3-1 through 3-13 are provided for comparison and a detailed calculation based upon actual S/d proximity factors should be performed to ensure that the IWB-3500 inspection standards are satisfied.

Flaws detected during inspection must satisfy the requirements of IWB-3500 standards to justify continued operation. However, if a flaw does not satisfy these ]

requirements, additional analysis may be perfonned in accordance with IWB-3600, which I allows for the use of analytical procedures to evaluate flaws tojustify continued operation.

i 1

25

L GENuclear Energy GENE-323-008-0194 Rev.1 DRF # 137-0010-7 l

000, M6Dd8-M ATTACHMENT LI PAGE 2_ / OF M 3.1.2 IWB-3600 Analytical Acceptance Criteria Per IWB-3600 [Ref.1], the analytical techniques described in Section 2.3 of this report can be used to establish acceptance criteria for flaws which may not necessarily j satisfy IWB-3500 acceptance standards.

l To justify continued operation, fatigue crack growth and irradiation embrittlement with time were considered. These effects have been evaluated for one operating cycle (18 months) beyond the current 16 EFPY, as well as for end oflife. These allowances were used to adjust the allowables calculated per Section 2.3.4 above to establish a flaw acceptance criteria, such that, if a flaw satisfies the acceptance criteria, continued operation

, isjustified for the terms specified above.

! L 3.2 Development of Acceptance Diagrams The results from Section 2.3 are used in conjunction with the above evaluation to develop flaw acceptance diagrams showing allowable flaw depth (a or 2a) versus aspect ratio (a/L) for each flaw orientation and location. The limiting flaw depth per IWB-3600 is bounded by either Kallow or the t1/3 limit, which ever is more limiting. These limiting values are reduced by an amount Aafcg for surface flaws and 2Aafcg .for subsurface flaws !

to account for fatigue crack growth to compute the IWB-3600 acceptance standard as follows: !

i aallow = a- Aafcg (surface flaw) (3-la) or 2aa llow = 2a-2Aafcg (subsurface flaw) (3-lb)

\

For some cases, allowable flaw sizes were found to be relatively small for the minimum specified test temperatures. But since the hydrotest condition is limiting for vessel weld regions, the allowable flaw sizes can be increased by increasing the test temperature. Therefore, to ensure added margin for flaw acceptability, the system hydrotest may be performed at higher temperatures.

26

GENuclear Energy GENE-523 008-0194 Rev.1 DRF # 137-0010-7 Doc a OEPd98 oM ATTACHMENT b4 PAGE 2N OF W Flaw acceptance diagrams for postulated axial and circumferential flaws at all selected locations are shown in Figures 3-1 through 3-13. Flaw orientation shall be classified as circumferential if the plane of the flaw is within 30 of horizontal. A flaw which is greater than 30* from horizontal shall be classified as axial. The curves are limiting for all normal, upset, emergency, and faulted operating conditions. During operation, the metal service temperature will be on the order of 400-550'F, which will always yield fracture toughness values higher than those computed for hydrotest or LOFWP.

Additional calculations were performed for specific weld areas in the beltline region. The maximum fluence level in the beltline region was adjusted for these location-specific (i.e. less conservative) analyses, which took advantage of the fluence levels fluctuating with respect to weld elevation and azimuth locations (Ref. I1). The allowable flaw curves are based on fluence levels shown in Appendix A, Table A-3.

Figure 311 shows the allowable surface flaw sizes for welds Vla, Vib, and Vic, which are located along the vessel azimuths 18',138*, and 258', respectively. Only the fluence at the top portion of the "VI"-welds, rather than the peak fluence, was used in the analysis (Figure 1-1). Figure 3-12 shows the allowable subsurface flaw sizes for the "Vl"-welds. Figure 3-9 shows the allowable surface flaw sizes for the "V2"-family of welds (Figure 1-1). Weld V2a is located at 60' azimuth, V2b at 180', and V2c at 300'. I Figure 3-10 shows allowable subsurface flaw sizes for the "V2 "-welds. I The use of Figures 3-1 through 3-13 is described in further detail in Appendix B, where a flaw evaluation procedure, flowchart, and worksheet are provided. !

27

[ __ _ _ _ _ --

GENuclear Energy GENE-323-008-0194 Rev.1 DRF # 137-0010-7 l

l ooc , _LIF.D4%-443

l. ATTACHMENT 3.I l

FAGE 24 0F - 75 3.3 Relative Ranking of Weld Regions Based on the allowable flaw sizes for an aspect ratio of a/L=l/6, each weld location has been ranked in Table 3-1. For any given weld region, the minimum flaw depth is determined regardless of flaw classification (i.e. surface or subsurface) or flaw orientation.

However, it was found that surface flaws were limiting in all cases. The rankings of the welds are then established based upon comparative rankings of these minimum flaw sizes.

As expected, the beltline region welds (VI, V2, and H12) also rank high because of irradiation embrittlement effects. The rankings in Table 3-1 may be used to prioritize planned inspections.

28 l

GENuclear Energy GENE-523 008-Ol94 Rev.1 DRF M 137 0010-7 DOC

M D 9 8 -d 4 8 ATTACHMENT 3*I PAGE A O_ OF 7F l

TABLE 3-1 Relative Ranking of Weld Regions Based on Allowable Flaw Sizesa for Aspect Ratio of a/L=l/6 = 0.17 Surface Flaws l

(

Ranking Weld Region Flaw Depth [in] Flaw Type i (See Fig. I) I (incl. clad) i 1

1 V2b o,450c axial

)

2 VI b o,450c axial 3 H12 b 0.450c axial 4 VFW inside 0.450c circumferential 5 VFW outside 0.450c circumferential 6 V3 0.865 axial 7 V4 1.073 axial 8 H23 1.316 axial 9 H34 1.590 axial i

aAllowable flaw sizes can be increased by increasing the hydrostatic test temperatures in raost cases.

bBeltline welds cIWB-3500 evaluation limiting - see Section 3.1.1, 29

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GENuclear Energy GENE %23-008-0194Rev.)

JRF M 137-0010-7 00C o NM%&

ATTACHMENT 5.I 4.0

SUMMARY

' PAGE - M OF ~7 6 A structural flaw evaluation was performed in accordance with ASME Code Section XI (1989 edition) for axial (meridional or longitudinal) and circumferential welds in the vessel shell regions. The analysis assumes the most limiting loadings for test, normal, emergency, upset, and faulted operation. The flaw evaluation pmvided in this i report includes fatigue crack growth and irradiation embrittlement for up to one operating cycle beyond the current 16 effective full power years (EFPY), and for end oflife (EOL) operation. In general, inside surface flaws were found to be limiting for vessel shell welds.

Evaluations were also perfonned for subsurface flaws for all selected weld regions.

Figures 1-1 and 1-2 of Section 1.0 identify all the weld regions selected for analysis.

Loading was assumed to be due to:

Membrane pressure stresses

Weld residual bending stresses

Clad residual stress (clad thickness = 5/16 in. nominal) on the inside surface only.

i Preliminary analysis found that hydrotest or LOFWP conditions were limiting for all welds because higher temperatures at operating conditions have associated higher fracture toughness that more than offsets increased stresses due to thermal transients.

Thy, only hydrotest and LOFWP conditions were further considered for fracture analysis. !

dasdd,ca the P-T curve for 16 EFPY, the minimum specified hydrotest temperature is 195 F at 1100 psig.

The analysis methods follow those prescribed in ASME Code Section XI IWB-3600. Applied stress intensity factors, K ,I were developed as a function of the flaw depth ratio, a/t (surface flaw) or 2 alt (subsurface), and aspect ratio, a/L. These were compared to the allowable fracture toughness, Kl a, reduced by the Section XI safety factor of 410 for !

hydrotest, to determine allowable flaw sizes.

l 43

q GE Nuclear Energy GENE-323-008-0194 Rev.I DRF G 137-0010-7 000 # NEM-dM ATTACHMENT Ll PAGE 4b 0F US~

An upper bound on allowable flaw size was established at 1/3 depth of the LAS wall thickness to ensure that ASME Code Section III primary stress requirements were met. A lower bound for allowable flaw sizes is established by the minimum inspection standards of IWB-3500. If the flaw, however, does not satisfy this standard, continued operation may still be justified if the flaw satisfies the IWB-3600 acceptance criteria, as I developed in this report. For the latter case, there is a reinspection requirement imposed by the Code (see ASME Section IWB-3132).

The allowable flaw curves presented in this report are base:1 on conservative assumptions of possible loadings and flaw location. If a specific flaw were to be found, )

which did not meet the IWB-3600 allowable in this report, a more specific analysis of the l flaw may show continued operation to be acceptable. If flaw-specific analysis to IWB-3600 criteria were not met, it would likely be possible to show continued operation to be acceptable on the condition that the pressure test temperature be increased. In an extreme case, either. flaw removal or weld repair might be necessary (see ASME Section IWA-4000).

1 i

l 44

- ~--

GENuclear Energy GENE-323-008-0194 Rev.1 DRF M 137-0010-7 Doc J A J ., g

5.0 REFERENCES

ATTACWENT J, g

~

PAGE OF

[1] ASME Boiler and Pressure Vessel Code,Section XI,1989 Edition.

[2] Swift, T.J., " Structural Evaluation of Head Flange Weld, Vessel Flange Weld and Upper Shell Weld at Quad Cities Nuclear Power Station, Unit 2", GE Nuclear Energy, SASR 90-82, December 1990

[3] Caine, T.A., " Cooper Nuclear Station Reactor Pressure Vessel Surveillance Materials Testing and Fracture Toughness Analysis", GE Report No. MDE-103-0986, May 1987.

[4] Rybicki, Shadley, Sandhu, and Stonesifer, " Experimental and Computational Residual Stress Evaluation of a Weld Clad Plate and Machined Test Specimens",

Journal of Engineering Materials and Technology Vol. I10, October 1988, pp.297-304.

[5] Ganta, Ayres, and Hijeck, " Cladding Stresses in a Pressurized Water Reactor Vessel Following Application of the Stainless Steel Cladding, Heat Treatment and Initial Service", PVP-Vol. 213/MPC-Vol.32, Pressure Vessel Integrity, ASME 1991.

[6] Ferrill, D.A., et. al., " Measurement of Residual Stresses in Heavy Weldment",

Welding Journal Research Supplement Vol. 45, November 1966.

[7] Landerman, E., and Grotke, G., " Residual Stress Considerations in Weldments for the Nuclear Industry," Weldments: Physical Metallurgy and Failure Phenomena, General Electric,1977.

[8] ASME Boiler and Pressure Vessel Code,Section III.1989 Edition.

[9] US Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, May 1988.

45

DOC o N EDf. 9 8-dM '

^

GENuclear Energy w

~' * GENE-523-008-0194 Rev.1 DRF M 137-0010-7 (10] Paris and Sih, " Stress Analysis of Cracks", ASTM Special Technical Publication No. 381, Fracture Toughness Testine and Its Anolications.1965.

[11] Caine, T.A., " Cooper Nuclear Station Vessel Surveillance Materials Testing and Fracture Toughness Analysis", GE Report No. 523-159-1292, February 1993.

[12] Letter from T.A. Caine to M. Bennett: " Review of Current Cooper P-T Curves Against Methods of Reg. Guide 1.99, Revision 2", May 18,1992.

[13] Lai, Wilson, " Finite Element Analysis of LOFWP Transient Effect on Vessel Wall," part of DRF-137-0010-7, October 1998.

[14] DeSalvo, G.J. and Gorman, R.W., ANSYS Engineerine Analysis System User's Manual. Revision 5.3, Swanson Analysis Systems, Inc., Houston, PA, October, 1996.

)

I

\

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l. i e 4

GENuclear Energy GENE-523-008-0194 Rev.)

DRF H 137 0010 7 i

I coc , MG Dc96-448 )

ATTACHM[NT 3-I PAGE M OF 78 f l

i i

I i

l i

APPENDAX A - Materials Data Book and Calculations j l

i i

l 47 m

GENuclear Energy GENE-523-008-0194 Rev.)

DRF H 137 0010-7 000 # U E Dd' M - d i b ATTACHMENT %+l A1. Vessel Geometry pagg 49 g, 75 Vessel dimensions at each weld location are listed in Table A-1. The clad thickness in the vessel region is 5/16"=0.3125 inch nominally. Figures 1-1 and 1-2 of Section 1.0 identify the location of the various shell plates and the welds. A LAS wall thickness of 5.375 inches was used in all of the cylindrical shell locations, except at weld VI, which '

was analyzed using a LAS wall thickness of 6.375 inches.

l A2. LimitingInitial RTndt The initial RTndt values for the vessel shell plates and welds are summarized in Reference [12]. For each weld locati on, the most limiting RTndt of either the weld or !

adjacent shell materials was used. The limiting initial RTndt values used in the analysis are listed in Table A-2.

A3. Limiting RTndt or f Non-Beltline Regions 1

Irradiation embrittlement is insignificant for non-beltline regions. Therefore, any shift in initial RTndt is negligible.

A4. Adjusted Reference Temperatures (ART) for Beltline Regions The allowable fracture toughness, K la, was computed for beltline regions based upon the adjusted reference temperatures computed in this section.

The ART values (including irradiation shifts and margins) are computed per US l

NRC Regulatory Guide 1.99 [Ref. 9]. The more limiting ART of either the shell (course) or the weld material was used in the analysis. l A4.1 Chemistry Data Based upon the Reg. Guide 1.99 methodology, the limiting chemistry factors, CF, I were computed for each beltline weld and are shown in Table A-3.

, 48

GENuclear Energy GENE-323-008-0194 Rev.)

DRF # 137-0010-7 00C a NEIE98-048 AT TACHMENT -- EI PAGE- b D OF ~7 9 A4.2 FluenceInformation Based upon a surface fluence at 16 EFPY [Ref.12), the neutron fluence, f[1019 nyt

( > 1 MeV)], at any depth in the vessel wall can be calculated by, f(x) = fsurf(e-0.24x) (A-1) where,

=

fsurf calculated value of the neutron fluence at the inner wetted surface of the vessel at 16 EFPY.

x (in) = depth into the vessel wall measured from the inner wetted surface.

Surface fluence information is listed for comparison in Table A-3.

Location-specific analyses were performed for welds in the beltline region. These analyses took into consideration the varying fluence levels, which fluctuated with respect to weld elevation and azimuth locations (Ref. I1). The assumed relative fluence variations were based on analyses of the reactor vessel. The allowable surface and subsurface flaw size curves for the "Vl" and "V2" families of welds are shown in Figures 3-9 through 3-12. These curves were based on fluence levels in Appendix A, Table A-3, as determined from the following equation:

fsurf(adj.)= fsurf x fluencerel-azimuth' x fluencerel-elev (A-1a)

! where, i

fluencerel-azimuth = normalized correction factor based on weld azimuth location

= normalized correction factor based on fluencerel-elev weld elevation 49

GENuclear Energy GENE-521-008-0194 Rev.I ;

DRF # 137-0010-7 \

occ , _ M EDIA19-646 ATTACIWM E(

PAGE - 9I 0F -- W .

(

The axial or vertical seam welds Vla, Vib, and Vic, are located at vessel azimuths !

18*,138', and 258*, respectively. The azimuthal locations of these welds were used to I'

obtain relative fluence levels based on information provided by Reference 11. A relative fluence level based on elevation was also determined (Ref. I1). Only the fluence at the top of the "Vl"-welds, rather than the peak fluence in the beltline region, was used in the analysis. A height of 44.125 inches above bottom of active fuel (BAF) was selected for determining relative fluence with respect to elevation; the region of the "Vl"-welds at this height is exposed to the highest possible fluence level for these welds. Fluence

)

. information for the "V1 " welds is shown in Table A-3.

Figures 1-1 and 1-2 show the "V2"-family of welds. Weld V2a is located at 60' azimuth, V2b at 180', and V2c at 300 . Again, relative fluence levels based on azimuthal locations were obtained from Reference 11. As shown in Figure 1-1, much of the "V2" welds is located in the beltline region. Table A-3 summarizes the fluence levels considered

'in the evaluation of the "V2" welds.

l A4.3 Calculation of ARTndt The irradiation shift in reference temperature, ARTndt, is the mean value of the adjustment in reference temperature due to irradiation embrittlement and is calculated per US NRC Reg. Guide 1.99 as follows:

ARTndt = (CF) f(0.28 - 0.10 log f) (A-2) l where, f= fluence [1019 nyt ( > 1 MeV)] at a specified depth and EFPY, i

l l

, 50 I

1

)

GENuclear Energy GENE-323-008-01N Rev.!

DRF # 137-0010-7 coe , _ MEo496-446 ATTACHMENT 3*I PAGE N OF W A4.4 Margin l

A margin term is included in the calculation of the ART to ensure a conservative !

upper bound. The values are listed in Table A-3 and are computed per US NRC Reg.

Guide 1.99 as follows:

Margin = 2 V(oI 2 + og2) I (A-3) where, oi (standard deviation ofinitial RTndt ) = 0'F

=

oA (standard deviation of ARTndt) The lower of 28'F or 1/2 of the mean RTndt value of the material- Weld Material

=

The lower of 17*F or 1/2 of the mean RTndt value of the material- Base Material i A4.5 Calculation of ARTS i

ARTS are calculated as a function of crack depth for each weld material, metal temperature, and EFPY per US NRC Reg. Guide 1.99 as follows:

ART = Init. RTndt + ARTndt + Margin (A-4)

Since ARTS are a function of crack depth, only values evaluated at the surface are listed in Table A-3 for comparison.

4 1

, 51

.]

GENuciaar Emrgy GENE-523-008-0194 Rev.)

DRF M 137-0010-7 DOC # NWWAbY ATTACHMENT 'E = l AS. Fracture Toughness,KIa paas M or W i Fracture toughness values are calculated per Appendix G of Section XI from the ASME Code [Ref.1] using the following expression, _

K Ia = 26.78 + 1.233e[0.0145(T-RTndt+160)] ksiVin (A-5) where, T( F) = metal service temperature (i.e. test temperature)

For beltline regions, the ART value is substituted for the RTndt. Fracture toughness values are sununarized in Table A4.

1 A6. Vessel Test Pressures and Metal Temperatures L

1 To determine the vessel metal temperatures, the P-T diagram from the technical specifications report for pressure-temperature limits [Ref.12] was used. For a pressure of 1100 psig during a hydro test, the minimum allowable metal temperature in the beltline region is 195"F at 16 EFPY.

A7. Bolt Preloed Stresses Boltup stresses are assumed negligible for the beltline vessel welds since stresses i

are fully attenuated. Beyond a 10 inch region of the flange discontinuity, flange stresses have been shown to be fully attenuated [Ref.2), and as such, the ' analysis for locations other than VFW does not include any boltup stresses.

. A8. Cladding ResidualStresses A stainless steel (SS) clad was applied to the inner surface of the vessel. Based on the simulated post-weld heat treatment (PWHT) data reported for the welds, a P%HT temperature of 1150 'F is assumed. For the purposes of this analysis, the cladding is assumed to have zero stress at i150 F. At temperatures other than the PWHT temperature, I

q

, $2

g. ,

00C , M E D 4 9 8 -4 4 8 GENuclear Energy __ ATTAchulut 3* , gggg_333_ggg_g,94 5,, ,

PAGE "" * "

DRF M 137-0010-7 clad stresses are difficult to predict, since they depend on the PWHT and the clad yield strength.

l For room temperature conditions the elastic clad residual stress is initially estimated at o c,70'F = E70(An)(AT)/(1-p) = 39.3 ksi > S y,70ep where, E70

=

28,300 ksi, modulus of elasticity for SS at room

{

temperature (Table i I-6.0,Section III of ASME Code [Ref. 8]).

= 0.3, Poisson's ratio.

f AT =

(1150-70) = 1080*F, I Aa =

Difference in coefficient of thermal expansion between SS and LAS,0.9 x10-6 in/in *F [Ref. 5) !

Sy,70op = 35 ksi, assumed SS yield strength.

However, because the clastic stress is greater than reported yield strengths for SS clad, a clad stress of 35 ksi will be used at room temperature. This value is consistent with typical yield strength values reported for 304SS clad materials.

The clad residual stress for the hydrotest condition at 195 F (for all regions excluding the bottom head) is estimated as, I

i oc,195 F " o ,70 c F + E195 (An)(AT)/(1-p) != 24 ksi where, i

E195

27,600 ksi modulus at 195 F AT

(70-195) = -125"F (hydro @ l95 F)

Aa =

2.54 x10-6 in/in- F '

=

oc,70 F Sy,70op = 35 ksi

. 53

1 GENuclear Energy GENE-323-008-0194 Rev.) i DRF G 137-0010-7 Doc , JJE.Dc 98-o4g Al TACHMENT _ .3. I PAGE b. E._of___ 7$

TABLE A-1 Vessel Geometry Weld Course # Maximum Min. LAS Clad Distance Region I.D. Thickness from flange

[in] [in] [in] [in]

Vla,Vib,Vic Lower 221 6.375 0.3125 >10 H12 Lower 221 5.375 0.3125 >10 V2a,V2b,V2c Lower-Inter. 221 5.375 0.3125 >10 H23 Lower-Inter. 221 5.375 0.3125 >10 V3 Upper-Inter. 221 5.375 0.3125 >10 H34 Upper-Inter. 221 5.375 0.3125 >10 V4 Upper 221 5.375 0.3125 >10 VFW Ves. Fig. 221 5.375 0.3125 =2 NOTE: For welds >10 inches away from the flange, boltup stresses are fully attenuated (Ref. 2).

. 54

GENuclear Energy GENE-523-008-0194 Rev.1 DRF # 137-0010-7 000 a Mc98 o4A ATTACHMENT _ _4,j PAGE e5g$ OF - 7G TABLE A-2 Limiting Initial RTndt Weld Course # Weld Matl.* Base Matl."

Region [*F] [ F]

Vla,Vib,Vic Lower -50.0 14.0 (beltline)

H12 Lower -50.0 14.0 (beltline)

V2a,V2b,V2c Lower-Inter. -50.0 14.0 (beltline)

H23 Lower-Inter. 0.0 10.0 V3a,V3b,V3e Upper-Inter. 0.0 10.0 H34 . Upper-Inter. 0.0 10.0 V4a,V4b,V4e Upper 0.0 10.0 VFW Ves. Fig. 0.0 10.0

Conservatively assumed zero for non-beltline welds

Based on limiting values of adjacent shell materials i

I l

u

-)

1 l

i GENuclear Energy '

GENE-523-008-0194 Rev.I DRF H 137-0010-7 Doc e _Mh98-04A TABLE A-3 Adjusted Reference Temperature (ART) Information

['"$k a 5 at Surface for Beltline Region @ 16 EFPY Weld fsurf' CF' al cA ARTndt Init. RTndt Surface Region [n/cm2) [op) [opj [opj [oF] ART [*F]

Via 6.09x1017 153 28.0 0.0 14.0 14.0 70.0 Vib 1.20x1018 153 41.5 0.0 17.0 14.0 89.5 Vic 6.53x1017 153 29.2 0.0 14.6 14.0 72.4 H12 1.20x1018 153 41.5 0.0 17.0 14.0 89.5 V2a 8.84x1017 153 34.9 0.0 17.0 14.0 82.9 V2b 9.35x1017 153 36.0 0.0 17.0 14.0 84.0 V2c 8.84x1017 153 34.9 0.0 17.0 14.0 82.9

Fluence and chemistry factor values for base metal (limiting material) l 1

l

}

i-56

GENuclearEnergy GENE 523-008-0194 Rev.1 l DRF # 137-0010-7 l

l DOC e NfDLAb-6% l ATTACHMENT 1I PAGE 38 OF - 73 l TABLE A-4 Fracture Toughness, KIa per Appendix G, Sec XI of ASME Code ,

I (Vessel material yield strength = 47 ksi)

. Weld Region Test Temp Surf. ART kip l

[*F] [F] [ksivin] l VIa 195 70.0 103.64 Vib 195 89.5 84.71 V1c 195 72.4 101.01 H12 195 89.5 84.71 V2a 195 82.9 90.53 V2b 195 84.0 89.52 V2c 195 82.9 90.53 H23,H34 195 10 200.00*

. V3,V4 195 14 199.89 VFW 195 10 200.00* l

'Kla value govemed by ASME code material limitations l

. 57

7 l-

\ GE Nuclear Energy GENE-323-008-0194 Rev.I DRF # 137-0010-7 l

DOC e UEuc 9A-644 l

ATTACHMENT 1i PAGE M 0F- 76 TABLE A-5 Applied Stresses (Vessel material yield strength = 47 ksi)

Location Weld Orient Pressure Bolt Weld Clad Stress Preload Residual Residual

m *b

[ksi) [ksi] [ksi) [ksi]

Ves. Fig. VFW axial 22.6 2.7 13.0 0 24 circum. I1.3 0 42.9 8 24 Ves. Shell VI, V2, axial 22.6 0 0 17 " 0**

V3' cimum. 11.3 0 0 9'* 0" j i

V4* axial 22.6 0 0 8 24 circum. I1.3 0 0 0 24 H12,H23 axial 22.6 0 0 9" 0" circum. 11.3 0 0 17** 0**

H34 axial 22.6 0 0 0 24 circum. 11.3 0 0 8 24 l

Note: It cludes the vertical" family" of welds. For example, weld VI includes Vla, Vib, and Vic

- LOFWP condition more limiting for welds in the middle to lower vessel region i

58

GENuclearEnergy GENE-323-008-0194 Rev.I DRF H 137-0010-7 DOC a NONd-48 ATTACHMENT - 3,i PAGE E _ 0F _ 78 APPENDlX B - Flaw Evaluation Procedure 59

.q l

I GENuclear Energy GENE-323-008-0194 Rev.) \

DRF M 137 0010-7 00c o MLDL.98-644 B. FLAW EVALUATION PROCEDURE ~ ^"^C "" 3

FAGE bI 0F 75 L This section describes the procedure to be followed to evaluate a flaw should one be found in the reactor pressure vessel. The attached worksheet is to be used to record the i flaw evaluation for each flaw. Figure B-1 may be used to sketch the flaw. Figure B-2 is a flowchart which outlines the detail evaluation procedure described below.

l The following procedure should be used in conjunction with the flaw acceptance

! diagrams (Figures 3-1 to 3-13), the attached worksheet and flaw evaluation flowchart to determine fiaw sizes and evaluate the acceptability of flaws.

Bl. Determine Region and Orientation of Flaw l

I Particular flaw locations not considered in this handbook are listed below. Flaws in any of these locations shall be listed as region 0 in step 1 of the worksheet. These flaws l must be addressed with a flaw specific analysis not covered by this handbook. '

a) Flaws in or near (within 1 plate thickness of) attachment welds, b) Flaws in or near (within VRt = 28 inches of) the shroud support plate to vessel weld, c) Flaws in or near (within JRt = 28 inches of) the vessel support skirt to vessel weld, d) Flaws in vessel studs or nuts, e) Flaws in nozzles or nozzle-to-vessel blend radii, f) Flaws in reactorinternals, t g) Flaws in nozzle safe ends or piping, h) Flaws in vessel tophead or bottomhead.

The various vessel wele considered in this handbook are shown on Figure 1-1.

! Flaws should be classified as either inside surface, outside surface, or subsurface per the proximity rules of Section XI of the ASME Code. Note that outside surface allowable flaw size is determined only for the VFW weld region. All other outside surface flaws that are detected will require location-specific analyses. The correct weld region identification (e.g. VI, V2, H12) is to be recorded on step 1 of the worksheet. If the flaw is contained in two regions or if it is not possible to definitely determine in which region the flaw is located, the flaw is to be evaluated against the acceptance criteria for the more limiting region.

, 60

p GENuclear Energy GENE-323-008-Ol94 Rev.)

l DRF G 137-0010-7 00C a O AT TA0hMENT b*I PAGE = 0F M The flaw orientation shall be classified as circumferential if the plane of the flaw is within 30' of horizontal. A flaw which is greater than 30* from horizontal shall be classified as axial.

I B2. Flaw Geometry and Classification l J

i The geometry of the flaw or group of flaws in close proximity are first to be 1 sketched. This sketch should be attacled to the worksheet. Figure B-1 may be used for this sketch which shall irclude:

a) The measureu thickness, 't', of the low alloy steel vessel wall in the region

}

containing the flaw. If the measured thickness is not available, the design thickness from Table A-1 shall be used.

b) The measured clad thickness,'tclad', in the region of the flaw. If the clad thickness can not be determined, the nominal clad design thickness of 0.3125 inch is to be used per paragraph IWA-3320 of Section XI of the ASME Code, c) The location of the flaw with respect to the surface and to other flaws is to be sketched and dimensioned in accordance with IWA-3300 of Section XI, ASME Code.

d) The flaw shape and measurements of proximity parameters in accordance with the proximity mies ofIWA-3300 of Section XI, ASME Code.

e) Combine any flaws in close proximity to other flaws or to the surface according to the rules cantained in section IWA-3300. Flaws need not be combined unless they are in parallel planes within 1/2-inch of each other.

pon applying proximity rules, planar flaws will be classified as follows:

1)inside . Jface flaws (measured from clad surface).

2)outside surface flaws (no clad).

3) subsurface flaws.

The appropriate classificat.' hould be recorded in step 3 of the worksheet.

61

p.

GENuclear Energy GENE-523-008-0194 Rev.!

DRF M 137-0010-7 boc a d f P d 8 M B3. Flaw Size and Aspect Ratio f,'g*,"U og 5 The final flaw dimensions are next recorded in step 5 of the worksheet. Note that for subsurface flaws, the flaw depth is measured as 2a. The flaw aspect ratio is calculated as a/L, where 'a' represents the half-depth in the case of subsurface' flaws.

B4. IWB-3500 Flaw Evaluation The detected flaw is first evaluated in accordance with paragraph IWB-3500 of the ASME Code,Section XI. Note that the subsurface IWB-3500 curves have been calculated only for a typical mid-plane flaw with a proximity factor of Y=1.0. The allowable flaw depth should be calculated in accordance with IWB-3510 based upon the actual measured flaw location to show acceptability. The subsurface IWB-3500 evaluation provided in this handbook are only valid for Y=S/d=1.0.

B5. IWB-3600 Flaw Evaluation l

If the flaw does not satisfy the requirements of IWB-3500, the IWB-3600 j allowables may be used to justify continued operation based on the analysis /results of this handbook. Reinspection, however, will be required (see ASME Section IWB-3132). 1

! L l

B6. Section III Evaluation (1/3 Limit) l If the flaw does not satisfy the allowable'per IWB-3600 analysis, either flaw removal (1/3 limit satisfied) or weld repair (1/3 limit exceeded) will be necessary, unless a further flaw-specific evaluation can show the flaw to be acceptable. Refer to Figure B-2 and the attached worksheet for further details, s

62 l-

GENuclearEnergy GENE-523-008-0194 Rev.1 DRF # 137-0010-7 I

f 00:, A)fd;M9e 448 i ATTACHMENT _ 3,I PAGE. b d 0F_75 B7. Further Evaluation i

If a flaw can'not be shown to be acceptable according to the flaw acceptance diagrams given in this report, it is possible that a flaw-specific analysis can be completed to show acceptance of this flaw with no repair. Because this analysis considers most of the Cooper Nuclear Station vessel, some conservative assumptions were made to make the re m en e 0 t sis I de:

1) The most limiting (highest) RTndt for any material within the adjacent shell segments is assumed for the entire weld region.

2) The largest bending stresses for the closure flange regions are assumed to occur over the entire region. ;

3) Subsurface flaws are evaluated based upon surface fluence values and bounding proximity effects.

Although these assumptions do not add much conservatism to the analysis in most l cases,'a flaw-specific analysis should first be conducted to eliminate these conservatisms.

Another possible alternative is to increase the hydrotest temperature. This will increase the material toughness during the most severe loading in terms of fracture and increase the allowable flaw depths.

If the local stress limit (1/3 thickness) region of the acceptance curve is exceeded (the flat portion of the curve) the assumptions made above will not affect t1 e flaw depth limit. For this case, a finite element analysis will likely be able to show additional maigin.

If a fhw-specific analysis is not able to resolve the detected flaw, the flaw may have to be ground out or in extreme situations weld repaired (see flowchart of Figure B-2).

I I

. 63 L

GENuclear EnerD' GENE.523-008-0194 Rev.) i DRF H 137 0010 7 DOC a SED 4.46-44f 1

l Vessel Flaw Sketch AMACHW 3, l '

1 PAGE b9 0F ---. 79~

Flaw ID:

)

l 4

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,h

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, g , D g i

g g i

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Cladding Thickness Vessel Wall Thickness t4 - , t=

LAS FIGURE B.I. Form for Vessel Flaw Sketches.

64

GENuclear Energ GENE-323-008-0194 Rev.1 l DRF # 137 0010-7 Doc

N 0 W W ~ $

FLAW EVALUATION FLOWCHART ATTACHMENT 3+I PAGE M OF 78

inerne - wow -

. s= newaoppi, - .r .

I e cinesew(antece.autenvece)

D -- w p-em.-e ce. u l

c= = = a e tece.e - en -

diegen YES Does w schafy IWB4500 creens?

l NO i f YES Does flew setsfy M64600 cntens?

fNO , '

YES is flew below tw 1/3 weg knt?

Y l NO ww.e- > Y y -

Wald Reper Requbed No fosispeckon

~

Y couTwueD OPERATION Flow Removal is e JUSTIFIED (no woid reper)

FIGURE B-2. Flowchart of flaw evaluation process 65

g GE Nucl:ar Energy DOC 0 N080~048 95y5 323 99g gjg4 p ,,,,

ATTAC m ENT %8 99p y j37 ggjg7 i

PAGE U7 0F 76

' COOPER NUCLEAR STATION FLAW EVALUATION WORKSHEET l FlawID:

1)' Dstgrmine Region and Orientation of Flaw. The weld region should be identified by the nearest weld. The orientation is either [A]xial or [C]ircumferential. If the flaw is at ajunction between two welds, the region with the more limiting acceptance criteria should be conservatively used.

Recion:

Orientation:

2) Sketch Flaw Geometry. Use the attached flaw sketch to draw the flaw.

I

3) Classify Flaw. Combine flaws in close proximity to other flaws and to the surface per the proximity rule ofIWA-3300,Section XI of the ASME Code. Classify s i

flaw as either:

Inside Surface Outside Surface Subsurface

4) Detennine Vessel Wall Geometry. If the flaw is classified as subsurface or outside surface, input 0 for clad thickness, else enter the analysis value for clad thickness as listed in Table A-1 of Appendix A for the specified wcld region. I Cladding Thickness, tclad =

(in)

Low Alloy Steel Thickness, tLAS =

(in) i Total thickness, t=tc lad + tLAS =

(in) 66

GENuclear L'nerg 000 0 M ~~

gym _. 15. ; GENE-523-008-0194 Rev.I DRF N 137-0010-7 PAGE M OF N (Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID:

5) Size Flaw. Calculate flaw depth, including any portion of the flaw extending into the cladding.

Surface Flaws: Subsurface Flaws:

Flaw Depth, a = (in) Flaw Depth,2a = (in)

Flaw Length, L = (in) HalfDepth, a = (in)

Flaw Length, L = (in)

6) Calculate Asoect Ratio of Flaw.

Flaw Aspect Ratio, a/L =

7) IWB-3500 Flaw Evaluatioa. For the given a/L aspect ratio, determine the allowable flaw depth, a (surface) and 2a (subsurface), in accordance with IWB-3510 of the Code and record the value below. If the flaw depth recorded in step 5 is below the allowable value, check the box " Acceptable per IWB-3500" below. Otherwise, check the box " Unacceptable per IWB-3500" and continue to step 8.

Inside Surface Flaw:

IWB-3500 Allowable Depth = a = (in) i Outside Surface Flaw (vessel flange regions only):

IWB-3500 Allowable Depth = a = (in)

Subsurface Flaw:

IWB-3500 Allowable Depth = 2a = (in) 67

Doce N E M -d M GENuclearEnergy ^

^

r GME423-00001Mi PAGE wJ OF ,

DRF # 137-0010-7 1

(Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID:

ACCEPTABILITY:

Acceptable per IWB-3500.

Unacceptable perIWB-3500.

8) IwB-3600 Flaw Evaluation. Record the appropriate flaw acceptance diagram Figure # from Section 3.0. Record the allowable flaw depth, a or 2a, from the appropriate curve for the specified orientation. If the flaw depth recorded in step 5 is below the allowable value, check the box " Acceptable per IWB-3600" below. Otherwise, check the box " Unacceptable per IWB-3600", and proceed to I step 9. 1 i

NOTE: Flaw specific analysis would be required if outside surface flaws were found in any region below the vessel flange. l Figure #

Inside Surface Flaw:

IWB-3600 Allowable Depth = a = (in) l Outside Surface Flaw (vessel flange regions only):

IWB-3600 Allowable Depth = a = (in)

Subsurface Flaw:

IWB-3600 Allowable Depth = 2a = (in)

ACCEPTABILITY:

Acceptable per IWB-3600. (for 16 EFPY)

Unacceptable per IWB-3600.

68

GENuclear Energy GENE-323-008-0194 Rev.1 Doc o DM 9A-6Ag DRF # 137-0010-7 AT TAChMENT _ 3.I PAGE - 70 _0F _ 76 (Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID:

9) From figure identified above, record the 1/3 wall thickness limit below. If flaw depth is below 1/3 limit, flaw removal is acceptable. Otherwise, weld repair is necessary.

1/3 Limit = (in)

From step 5 above:

Flaw depth = a = (surface) 2a = (subsurface)

Flaw excavation depth < 1/3 Limit: Flaw removal acceptable (No weld repair)

Flaw excavation depth > 1/3 Limit: Weld repair required I

69 l

i l

l 000 a_ LIE n g.9 g _ q ~

)

ATTACHMENT J, j ~

l PAGE ~7 / __ op .__ .-fy-S GE Outage Services l

October 27,1998 ,

l Mr. Michael Friedman l NPPD / Cooper Nuclear Station P.O. Box 98 Brownville, NE 68321

Reference:

GE/CNS98018

Dear Mike:

Attached is original signed version of the response letter generated by Wilson Lal, relative to the recent telephone conversations between Wilson and All Bacha. The additional fgures associated with this attached letter are meant to be supplemental only and are not to replace the fgures within the handbook.

If you have any questions or require any additiutal information please call me at extension 2874.

Sincerely, m /A-

./ / A Joseph M. Hewett til GE Project Manager Outage Services i

l l

l

]

1 Doc o AJEDe' 96--44f October 16,1998 ATTACHMENT - %i PAGE- 72. og _ 75

. Mr. Ali Bacha Cooper Nuclear Station Design Engineering Department i

l

Subject:

Cooper Nuclear Station i RPV Flaw Evaluation Handbook Rev.1 Report No. GENE 523-008-0194 (September 1998) i l!

Dear Mr. Bacha:

I l

This letter is prepared in response to our recent telephone conversation. The enclosed figures (Figure 3-14,3-15, and 3 16) show allowable sizes for outside flaws found in the beltline weld

' region, where the fluence level can be significantly different between the inside and outside surfa:e of the vessel wall. In general, allowable flaw sizes are more limiting for ID flaws than OD flaws (for the same weld); so the allowable flaw chrts currently shown in the RPV Flaw Evaluation !

Handbook may be used for inspecting both ID and OD flaws. Therefore, the additional figures provided with this letter are meant to supplement (not replace) those shown in the Handbook.

l This letter will be faxed to you. If you have any questions, please call me at 408-925-4011. Thank you.

Regards, AC Wilson Lai, GE enclosure: supplemental Figures 3 14,3 15, and 3-16 1

l

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NEDC 98-048 , Prepared By: General Electric (GE) Checked / Reviewed Dy: Ali Bacha Rev. No. O Date: 10/15 19 98 Date: 11/01 19 98 l

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ATTACHMENT 3.2 Structaral Integrity's Letters reference MLH-98-045, MLH-98-047, MLH-98-056, MLH-98-063, MLH-98-066, and MLH-98-067 l

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p cce. e Bws%4e ATTACHMENT - 37 PAGE-- 1 0F - //a i

August 21,1998 :

MLH-98-045 mherrera@structintcom l

l Mr. Michael J. Friedman Nebraska Public Power District l Cooper Nuclear Station P.O. Box 98 ,

Brownville, NE 68321

Subject:

Third Party Review of GE Flaw Acceptance Handbook:

GE Report GENE-523-008-0194 i

Dear Mike:

i l We have performed a review of the subject report. We have not completed a detailed evaluation 1

of the calculations since we did not have the references needed to complete this work. Per our phone call today, we expect to receive the references I requested on Monday. Our comments regarding the subject report are provided below.

Executive Summary Page 3 There is no justification provided in the report regarding the determination of limiting events. There is no detailed discussion of events other than boltup and hydrotest.

Page 4 Change " allowable fracture toughness" to " allowable stress intensity factor."

Page 5 Spell out all words.

Explain note c (IWB-3500 evaluatian is limiting).

Page 6 & 10 Labeling has been truncated in figure.

Page 7 Table of contents:" MATERIALS AND .." should be labeled as Appendix A.

i Page 8 Disposition time is significantly reduced.

Last sentence, what does this mean?

1 First bullet, were boltup stresses included?

I ;

i coc o MGDc4B-64A ATTACHMENT - L2 l t

PAGE- "2- 0F- M l Michael J. Friedman August 21,1998 l Page 2 r

MLH-98-045 Page11 Label top scale as " Azimuthal Location."

Label " Beltline Region."

l

. Page 12 1" paragraph, Boltup stresses?

Why weren't any of the thermal transients looked at such as LOFWP and i

shutdown cooling?

l Need detail and justification for assuming hydrotest is worst condition. There can l l be some thermal stress during some of the thermal events which would be possible based on the Cooper P-T curve. l l-Page 14 The weld residual stress was assumed to be bending. Does this mean tensile on ID and compressive on OD? The residual stress should follow a sinusoidal behavior, tensile on both ID and OD and compressive in the wall center. This may be non-conservative for OD flaws.

Why is the weld residual stress perpendicular to the weld assumed to be zero?

Page 16 Add bullet, " Vessel weld residual stress distribution and magnitude." i Do the eccentricity ratios take into consideration subsurface flaws on the ID half and OD half of the vessel wall?

Page 19 The K ranges may be non-conservative if there are some compressive thermal stresses present (during cooldown on OD).

Page 20 Any basis for 10 cycles per year?

Page 22 The figure shows the clad stress as distributed, but the text states that it was a point load at the clad mid point.

Page 27 1" paragraph, it is stated that curves are limiting for all normal and upset events.

What about other events, C & D?

Do these curves apply for both axial and circumferential flaws?

Do the surface flaw curves apply to both ID and OD flaws?

Was crack growth subtracted from the 1/3 limit curves?

Where are the curves for the different eccentricity ratios?

Page 47 Section A3 needs to be reworded.

Page 48 The exponent should be -0.24x, not -0.024x.

Fluence should be given as E>lMeV.

Do you mean the clad inner surface or clad /LAS interface? Are you taking credit

' for the attenuation through the clad thickness?

" fluencer a zimuth" is normalized correction factors not fluence level.

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occa G N 48-64 ATTACHMENT 3.1 PAGE E OF /b -

Michael J. Friedman )

August 21,1998 Page 3 MLH-98-045 ,

l l Page 50 RL should be mean RL.

Page 54 What is the basis for " conservatively assumed zero for non-beltline welds"?

Page 60 Per the ASME Code, both the vertical and horizontal component need to be i evaluated.

Page 67

)

Outside surface flaw (top head... Top head was not evaluated in this study. Also f

on Page 68.

If you have any questions, please do not hesitate to contact me. I would like to have a discussion with you early next week so we can proceed with completing the review.

Very truly yours, l

Marcos L. Herrera i Senior Consultant t

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DOC e- N h 98-048 ATTACHMENT ___ 12 PAGE 4 - Of - M August 25,1998 MLH-98-047 mherreragstu:bntco Mr. Michael J. Friedman Nebraska Public Power District '

Cooper Nuclear Station P.O. Box 98 Brownville, NE 68321

Subject:

Hird Party Review of GE Flaw AW== Handbook:

GE Report GENE-523-008-0194

DearMike:

We have performed a review of the subject report. Our comments regarding the subject report are pnwided below.

Eggytive Summary Page1 Signature blocks " prepared by","nnfied by", and " approved by" not signed Page 3 There is no justification provided in the report regarding the determination of limiting events. Dere is no detailed discussion of events other than boltup and hydrotest What is the reference for the P-T curve for 16EFPY?

Page 4 Change " allowable fracture toughness" to " allowable stress intensity factor."

Page5 Spell out all words.

Explain note c (IWB-3500 evaluation is limiting).

Page 6 & 10 Labeling has been truncated in figure.

Page 7 Table of contents:" MATERIALS AND..." should be labeled as Appendix A.

Page 8 Disposition time is significantly reduced.

Last sentence, what does this mean?

First bullet, were boltup stresses included?

Page11 Label top scale as " Azimuthal Location."

Label" Beltline Region."

Page 12 - l paragraph, Boltup stresses?

Why weren't any of the thermal transents looked at such as LOFWP and shutdown cooling?

Need detail and justification for assummg hydrotest is worst condition. There can

c .

Doc e NEV448 -648 ATTACHMENT "L 2 PAGE 6 0F /b be some thermal stress during some of the thermal events which would be possible based on the Cooper P-T curve.

Page 14 The weld residual stress was assumed to be bending. Does this mean tensile on ID and compressive on OD? The residual stress should follow a sinusoidal behavior, tensile on both ID and OD and compressive in the wall center. This may be non-conservative for OD flaws.

Why is the weld residual stress per-adielar to the weld assumed to be zero?

K.a should be consistent with K.n .

Page 16 Add bullet, " Vessel weld residual stress distribution and magnitude."

Do the eccentricity ratios take into consideration subsurface flaws on the ID half and OD half of the vesselwall?

Page 19 The K ranges may be non-conservative if there are some compressive thermal stresses present (during cooldown on OD).

Page 20 Any basis for 10 cycles per year?

Page 22 The figure shows the clad stress as distributed, but the text states that it was a i pointload at the clad mid point.

Page 25 & 63 S/d should be S/a.

1 Page 27 1" paragraph, it is stated that curves are limiting for all normal and upset events. )

What about other events, C & D?

Do these curves apply for both axial and circumferential flaws?

Do the surface flaw curves apply to both ID and OD flaws?

Was crack growth WW from the 1/3 limit curves?

Where are the curves for the different eccentricity ratios?

Last sentence, change to " procedure, flow chart, and wo:kshect" to reflect sequence laid out in handbook.

Page 34,35 36,37 Correct x and y axis and labels so they don't conflict. 1 Page 39 Figure 3-10 sixxild read " Cooper Subsurface "V2" Flaw-16EFPY" Page 42 Add " emergency" to considered limiting loadmg.

Page 43 Provide details of Code required reinspection.

Provide detailed Code reference for flaw removal or weld repair.

Page 47 Why use muumum wall thickness everywhere? Doesn't the nominal thickness vary at some locations up to 6.375". This may be very conservative.

Page 47,49 50 Peference for ART, CF, RTer calculations and provide copy for review.

Page 47 Section A3 needs to be reworded.

Page 48 The exponent should be -0.24x, not -0.024x.

Fluence should be given as E>lMcV.

Do you mean the clad inner surface or clad /LAS interface? Are you taking credit for he attenuation through the clad thickness?

"fluca~ 1" is normalized correction factors not fluence level.

n .-

DOC # -NED698-dd ATTACHMENT - 3,2 PAGE b 0F - /0 Page 50 RTa should be mean RTa.

Page 54 What is the basis for " conservatively assumed zero for non-beltline welds"?

Table A-2, V3 should read V3a, V3b, V3c. V4 should read V4a, V4b, V4c.

Page 60 Per the ASME Code, both the vertical and horizontal components need to be evaluated.

Page 62 What is a " Site Corrective Action Program Activity?"

Page 67 Outside surface flaw (top head... Top head was not evaluated in this study. Also on Page 68.

If you have any questions, please do not hesitate to contact me.

Very truly yours, Marcos L Herrera Senior Consultant S

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ooc a _MewAB-s4s StructuralIntegn.ty Associates, Inc. acaw u PAGE- 7 . OF -- /0 3315 Almaden Expressway Suite 24 September 15,1998 an a a cA E86 MLH-98-056 rI"' $$

mherrera@structintcom Mr. AliBacha Neliraska Public Power District Cooper Nuclear Station P.O. Box 98 Brownville,NE 68321 1

Subject:

Review of Cooper Nuclear Station RPV Flaw Evaluation Handbook Report No. GENE-523-008-0194, Rev. 0 (September 1998)

Dear Ali:

I have completed a review of the subject repon. Following are my comments on the repon. I have repeated my comments that I don't consider closed, followed by my reason. There are also some new comments that have been added.

Executin Summary General Comment: Ifhydrotest is the limiting condition, there should also be a statement that since the analysis covers all normal, upset, emergency andfaulted corulitions, it has thus also considered all allowable operating conditionsper the P-Tcurves.

Page 3 There is nojustification providedin the report regarding the determination oflimiting events. There is no detaileddiscussion ofevents other than boltup andhydrotest.

In many cases, the full hydrotest conditions are limiting (195 F,1100 psig), ;

but intermediate hydrotest conditions may also be limiting. For example, for j FitzPatrick, which is similar to Cooper, according to the P-T curve A, there are allowable operating conditions of120*F and 610 psi. Although the j pressure is lower, the temperature is well below that for the hydrotest. A !

second intermediate hydrotest condition is at 90 F and 312 psi along with a slight cooldown.

At FitzPatrick, we have also found that conditions along curve B could be limiting for outside surface flaws (212 F,1040psig, heat-up rate of100 F/hr.)

See Jose, CA . Ahres,0H Silver Spring, MD Pompone Beach, H Taipal,Taiwas Charlotte, NC Phors 408 978-8200 Phone: 330-864 8886 Phore 301589-2323 Phone: 954-917 2781 Phone: 02 388-5508 Ptane 704 573-1369 u

ooc o DEIM 99-64B ATTACHMENT _ ~3,2 PAGE - b _0F __ //a Mr. AliBacha September 15,1998 Page 2 MLH-98-056 The use of results from other plants to make assun'ptions is valid if both the strenes are similar and the material behavior is similar. In order to make the assumptions based on results from other plants, the RTuor. Guence, tc...would also need to be similar.

We have also found that sometimes the exclusion of the 8 ksi weld residual stress could lead to limiting conditions. For this reason, we perform calculations with and without weld residual stress. This is appropriate due to the variation of weld residual stress. The 8 ksi used in this malysis is likely a nominal value and may not necessarily be the limiting condition.

Page 6 & 10 Labeling he.s been truncatedinfigure.

There are still some truncated labels in the figure.

Page 7 Table of contents: "MA TERIALS AND... " should be labeled as Appendix A.

Appendix A and Appendix B are not consistent in Table of contents.

Page 12 Why weren 't cuny ofthe thermal transients looked at such as LOFWP and shutdown cochng?

During intermediate hydrotest conditions, the vessel may not be at isothermal conditions.

Need detail andjustificationfor assuming hydrotest is worst condition. There can be some thermal stress during some of the thermal events which would be possible basedon the Cooper P-Tcurve.

Same as previous comment.

Why is the weld residualstressperpendicular to the weldassumed to be zero?

For flaws perpendicular to the weld, the report states that the weld residual stress is not included. However, if a flaw is in the weld itself, regardless of orientation,it is subject to weld residual stress. I agree that the stress does decrease significantly, thus reducing the driving force for a long flaw, but that stress is still there for a shorter flaw. Since the flaw acceptance is based on stress intensity factor (LEFM) at the deepest point, I believe the stress should be included, f StructuralIntegrityAssociates,Inc.

Doc o AlILDc48-d44 ATTACHMENT L'2 1 PAGE 9 0F O Mr. Ali Bacha September 15,1998 Page 3 MLH-98-056 Do the eccentricity ratios take into consideration subsurfaceflaws on the ID half andOD halfofthe vesselwall?

It is very conservative to use the maximum eccentricity value.

Page 14 " maximum primary membrane stress" should be "marimum primary local stress".

Page 16 Comment only, no action: Although Individualstresses may be elastic, the net stress may be plastic and require more sophisticated evaluation.

l Page 17 You may want to state that Kesa isfor a longpaw.

Page 27 Firstparagraph, "normaland upset" shouldread " normal, upset, emergency, faulted".

l Page 43 Provide details ofCode requiredreinspection.

This was requested by the customer. I l

Provide detailed Code referenceforflaw removal or weld repair.

This was requested by the customer.

Page 47 Why use minimum wallthickness everywhere? Doesn't the nominalthickness vary at some locations up to 6.375". This may be very conservative.

Isn't there a weld at the transition between the 6.375" and 5.375" in the beltline? This does give different stresses and distributions than that for a constant thickness shell.

Flaw Diagrams: The mar allowable curves could be higherfor circumferentialpaws.

Was a Cooper specific stress analysisperformed to determine the bolt-up stresses using the qppropriate bolt loads, dimensions, sealloads, etc. This is unknown since the report does not contain any ofthese details.

Page 60 ' ' Theflaw should be broken into arial and circumferential components and analysed both ways.

Appendix B ODflaws were only evaluatedfor weld VFW, state so in Appendix B.

f StructuralIntegrilyAssociates,Inc.

000 0 MN 98-d48 ATTACHMENT 442 PAGE. IO Op I6 Mr. Ali Bacha September 15,1998 Page 4 MLH-98-056 Figure 3-2 " Site Correctin Action" was not deleted Figure 3-7 Allowable surfaceflaws are ginnfor V3, V4. The allowableforIWB-3500 is only 0.2"at att - O. Howewr, this depth is less than the clad thichwss of 0.3125". The allowable should be at least the clad thickness since in Section 3.1.1 it states that the allowable per IWB-3500 was added to the clad thichwss so that alldepths were with respect to the cladsurface.

Figure 3-10 x-arislabelneeds to be corrected Please contact me if you have any questions or want to discuss these comments further. -

Regards, k g ,n i arcos L. Herrera, P.E.

{

Senior Consultant 1 S

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f StructuralIntegrityAssociates,Inc.

ooce M EIE % 4A Structural Integnty Associates, Inc. mAe- sa PAGE d I 0F @

l 3315 Almaden Expressway Suite 24 October 6,1998 San Jon, CA 95H8G7 MLH-98-063

$"" $a-97s.,9 82 mherrera@stnictint.com Mr. Ali Bach.s Nebraska Pr'olic Power District Cooper Nuclear Station P. O. Box 98 Brownville,NE 68321

Subject:

Acceptance of Comment Resolution Regarding CNS RPV Flaw Evaluation Handbook

References:

1. Cooper Nuclear Station RPV Flaw Evaluation Handbook, Report No.

GENE 523-008-0194 (September 1998)

2. Phone Call, Wilson Lai, H. S. Mehta, M. L. Herra, Regarding Resolution of Comments,9/28/98. l

3. Letter, Joe Hewett (GE) to Ali Bacha, September 29,1998, Regarding Response to Comments

Dear Ali:

This letter is for the purpose of documenting the acceptance of the GE proposed resolution to SI comments regarding the Reference 1 report.

Based on the Reference 2 phone call and Reference 3 letter, I accept the responses as being sufficient to address the SI questions. The Reference 3 letter contains a statement regarding the hydrotest procedure which was agreed upon in the Reference 2 phone call. Although unlikely, operation in the unanalyzed regions of the P-T curve should be avoided unless analyzed in the future.

I would like to add that the handbook is conservative, very conservative in some cases (except intermediate hydrotest), and may result in subsequent refined analysis being required if flaws are detected.

NPPD may want to consider refining the analysis in the future in order to eliminate the potential

! need for emergency analysis that could impact the CNS outage.

l l

l Malase, CA Akten, OH Silver Spring, MD Pompano Beach, F1 Taipet Taiwan Chartone, IIC

] w 4 978-8200 Phons: 330-864-8886 Phone 301-58&2323 Phone:954 917-2781 Phone 02-388-5508 Phone: 704-573-1369

000 0 MWRB-64A ATTACHMENT & 2- i PAGE IA 0F Ib Mr. Ali Bacha October 6,1998 Page 2' MLH-98-063 1

I Please feel free to contact me if you have any questions.

Sincerely, i I

p_ /.,

Marcos L. Herrera, P.E.

)

l Senior Consultant li cc: A. F. Deardorff NPPD-09 f StructuralIntegrityAssociul:s,Inc.

L

.w i

ooc a MEntAB-o4e StructuralIntegrity Associates, Inc. ^nac~ u PAGE - I3 0F /[o 3315 Almaden Expressway suite 24 October 8,1998 san Jose, CA MM66 MLH-98-066 [," "' '$j@s2co mherreraestructint.com Mr. Ali Bacha Nebraska Public Power District Cooper Nuclear Station P. O. Box 98 Brownville, NE 68321

Subject:

Acceptance of CNS RPV Flaw Evaluation Handbook

References:

1. Cooper Nuclear Station RPV Flaw Evaluation Handbook, Report No.

GENE 523-008-0194, Revision 1, DRF #137-0010-7 (September 1998)

2. Phone Call, Wilson Lai, H. S. Mehta, M. L. Herrera,- Regarding Resolution of Comments,9/28/98.

3. Letter, Joe Hewett (GE) to Ali Bacha, September 29,1998, Regarding Response to Comments

Dear Ali:

This letter is for the purpose of documenting the acceptance of the GE RPV flaw evaluation handbook, Reference 1.

Based on the Reference 2 phone call and Reference 3 letter, I accept the responses as being sufficient to address the remaining SI questions. The Reference 3 letter contains a statement regarding the hydrotest procedure which was agreed upon in the Reference 2 phone call.

Although unlikely, operation in the unanalyzed regions of the P-T curve should be avoided unless analyzed in the future.

During the Reference 2 phone call, a review of all of the open comments was performed to assure that the handbook was technically sufficient. I would like to add that the handbook is conservative, very conservative in some cases (except intermediate hydrotest), and may result in subsequent refined analysis being required if flaws are detected.

NPPD may want to consider refining the analysis in the future in order to eliminate the potential need for emergency analysis that could impact the CNS outage.

Sea Jose, CA -

Akroa, OH Silver Spring, MD Pompase Beach, Fl Taipal, Taiwan Chartone L Phone. 408 978 8200 Phone. 330-864 8886 Phone:301 589-2323 Phone: 954 917-2781 Phone. 02 388 5508 Phone: 704-5731369

~

DOC OWW -

ATTACHMENT 3d2 PAGE I 4 -- 0F lA' Mr. Ali Bacha October 8,1998 Page 2 MLH-98-066 Please feel free to contact me if you have any questions.

Sincerely,

/

4, .~ -

arcos L. Herrera, P. E.

Senior Consultant l

( gsv l cc: A. F. Deardorff NPPD-09 l

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f StructuralIntegrityAssociates,Inc.

15/E3/S3 15:48 STRUCTtRAL INTEGRITY ASSOC INC o 4628253 93 NO.846 DB2 00c a d EDd%-M StructuralIntegrity Associates, Inc. arracuueur z.2 PAGE - N _ 0F __ Ib 3315 Aisnadat Expressway suite 24 October 8,1998 '"

MLH-98-067 Phone: 408-978-8200 Fax: 408-978-8964 rit ...geenntean Mr. AliBacha Nebraska Public Power District Cooper Nuclear Station P. O. Box 98 Brownville, NE 68321

Subject:

CNS RPV Flaw Evaluation Handbook Conservatisms and Possible Enhancem

References:

1.

Cooper Nuclear Station RPV Flaw Evaluation Handbook, Report No.

GENE 523-008-0194, Revision 1, September 1998.

2.

Letter, Joe Hewett (GE) to Ali Bacha, September 29,1998, Regarding Response to Comments

Dear Ali:

Per your request, I am providing some of the conservatisms or features that Nebraska Public Power District (NPPD) may want to consider evaluating in the future regarding the recently completed Reference 1 report.

1) Subsurface flaws are evaluated using maximum eccentricity ratios. The Reference 1 report uses a maximum eccentricity ratio (2 e/t) to determine the M. and M. factors for the stress connection. This is correct for flaws found near the surfaces, but is conservative for flaws away from the surfaces (closer towards the center of the RPV wall). A more realistic method would be to present allowable subsurface flaw dimensions for different values of 2e/t.

2)

Subsurface flaw locations (ID or OD). The locations ofsubsurface flaws (near inner surface or near outer surface, resulting in +/- 2e/t ratios) were not considered.

3) Sub.=wnce flaw proximity. Allowable flaw sizes are given for a proximity factor of Y =

S/a = 1.0 (typical for mid-plane flaw). IfY < 0.4, the flaw must be classified as a surface defect and would require reevaluation since this is not covered by the flaw handbook.

Evaluation should be performed with this check included. Thus, a 2e/t factor is calculated, and the allowable flaw is determined such that allowable Ki is exceeded or the flaw becomes a surface flaw (per ASME Code proximity check).

San Jena, CA Akrsa,0M Saver Spring. M0 Pomaano Osact, FI falpst Taiwan

~ . . . ~ . . . . . - .. .. .. Chanone. sic

10/08/S3 15:48 STRUCTURAL INTEGRITY ASSOC INC

4628255099 NO.846 903 Doc , _ UECV 9 '3-d48 ATTACHMENT 12 PAGE Ib 0F lb Mr. AliBacha October 8,1998

! Page 2 MLH-98-067 l

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4) Maximum allowable circumferential flaw. The 1/3 I . nit is applied to both circumferential and axial flaws. This is conservative for circumferential flaws since the pressure stress for axial flaws is twice that for circumferential flaws.

5) Outside surface flaws. The handbook hould ideally contain all possible locations and orientations including outside surface flaws at all locations.

6) Intermediate hydrotest conditions. Although the statement in the Reference 2 letter supports the current RPV inspection, there still may be a long-term concern of having regions of the P-T unanalyzed from the fracture perspective. NPPD may want to consider evaluating other key points during the hydrotest.

l Please contact me ifyou have any questions.

Sincerely, Y

/+

Marcos L. Herrera, P. E.

Senior Consultant P9 cc: A. F. Deardorff NPPD-09 l

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f StructuralIntegrityAssociales,Inc.

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i l H: t CNSPROCS t FORMS iVOL3 % 347 % 3+7.3 N1brasks Public Power District DESIGN CALCULATIONS SHEET sneet -

or - I NEDC 98-048 Prepared By: General Electric (GE) Checked / Reviewed By: Ali Bacha Rev.No. O Date: 10/15 19 98 Date: 11/01 19 98 i

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(

ATTACIIMENT 3.3 )-

GE's Comment Resolutions (Letters) l

I Doc e DPRRA-s4-8 ATTACHMENT EM PAGE I _ 0F II J

9 GE Outage Services l October 2,1998 l i

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Mr. Michael Friedman j

! NPPD / Cooper NuclearStation P.O. Box 98 Brownville, NE 68321

Reference:

GE/CNS98013

Dear Mike:

Attached is original signed version of Revision #1 of GENE -523-0084194 titled " Cooper Nuclear Station RPV Flaw Evaluation Handbook". Also attached is the cover letter addressing the comments I relative to the review conducted by Structuralintegrity. Based on the transmittal of this document and the completion of the RPV Flaw Handbook last week, this task is considered completed.

If you have any questions or require any additional information please call me at extension 2874.

Sincerely,

4. b Joseph M. Hewett 111 GE Project Manager Outage Services l

DOCe D N ATTACHVENT 3*3 September 29,1998 FAGE '2 - 0F II Mr. AliBacha CooperNuclear Station Design Engineering Department

Subject:

Cooper Nuclear Station RPV Flaw Evaluation Handbook Repon No. GENE-523-008-0194 (September 1998)

Dear Mr. Bacha:

Enclosed is the Rev. I copy of the flaw evaluation handbook. His letter a'so addresses comments documented in the e-mail letter dated September 15,1998," Review of Cooper Nuclear Station RPV Flaw Handbook; Report No. GENE-523-008-0194, Rev. 0 (September 1998)" from Marcos L. Herrera to Michael J. Friedman. Original comments are shown in Italics. Item by item responses are shown bold.

nese comments have already been resolved with Structural Integrity, and are provided here for your record. Please note in particulas the following (which is repeated in our enclosed responses):

Although the PT curves used for thk analysis allow for reactor operation in areas deemed more limiting than those used in our calculations (e.g. at 120*F and 610 psi, or at 90*F and 312 psi), these pressure / temperature conditions were not considered, because reactor operation is not expected to reach these zones on the PT curve. He flaw calculations used for this report are based on the expectation that Cooper reactor operation (i.e. power ramp-up and ramp-down) will never reach the limiting zones suggested above. Cooper documented procedures will need to be reviewed, and if they are inconsistont with this assumption, then GE will need to be notified immediately.

A final report will be submitted to you via e-mail, followed by a faxed copy of the signed cover sheet. A hardcopy of the report will be sent to Joe llewett at the following addmss:

Joe Hewett, GE c/o Cooper Nuclear Station 3.5 Miles off Route 136 Brownville, Nebraska 68321

. ph: 402-825-5018 If you have any questions, please call me at 408-925-4011. Thank you.

Regards, '

o* ,

I \

O j%

Wilson Lai, GE enclosure: response to comments Rev. I Flaw Evaluation Repon

[

00c e klFTE48-64A ATTACHMENT 2,3 PAGE 3 0F II Executive Summary. General Comment: Ifhydrotest is the limiting condition, there should also be a t statement that since the analysis covers all normal, upset, emergency andfaulted conditions, it has thus also consideredall allowable operating conditions per the P-Tcurves. l 1

l Done. A Loss of Feedwater Pump condition was also analyzed for vessel welds in the middle to lower I vessel refons, and this is mentioned in the report.

Page 3. There is najustopcation provided in the report regarding the determination oflimiting events. i There is no detaileddiscussion ofevents other than boltup and hydrotest. '

In many cases, thefull hydrotest conditions are limiting (195 *F, i100psig), but intermediate hydrotest conditions may also be limiting. For example,for FitzPatrick, which is similar to Cooper, according to the P-Tcurve A, there are allowable operating conditions of120* Fand 610 psi. Although thepressure is lower, the temperature is wilbelow thatfor the hydrotest. A secondintermediate hydrotest l condition is at 90

F and 312 psi along with a slight cooldown.

Although the intermediate hydrotest conditions mentioned above may be limiting, and the PT carves i allow reactor operation in those regions, actual plant operation is never expected to reach those limiting regions on the PT curve. Cooper start-up calls for low pressure operation with the l

I temperature steadily raised. Pressure is allowed to increase only after the required water I temperature is reached. Therefore, the vesselis never expected to experience the intermediate hydrotest conditions suggested above, i j

At FitzPatrick, we have alsofound that conditions along curve B could be limitingfor outside surfaceflaws (212* F,1040psig, heat-up rate of100* F/hr). 1 The use ofresultsfrom otherplants to make assumptions is valid ifboth the stresses are similar andthe materialbehavior is similar. In order to make the assumptions based on resultsfrom otherplants, the RTNDT, fluence, etc...would also need to be similar.

Only VFW outside surface flaws were evaluated for 1100 psig and 195 deg. F. All other detected outside surface flaws will require location-specific analysis.

We have alsofound that sometimes the exclusion ofthe 8 ksi weld residualstress couldlead to limiting conditions. For this reason, weperform calculations with and without widresidualstress. This is 1

appropriate due to the variation of wid residual stress. The 8 ksi used in this analysis is likely a nominal value and may not necessarily be the limiting condition.

1 For this analysis, stresses (bending or membrane) were applied as tensile stresses. Therafore, '

exclusion of the 8 ksi weld residual stress wculd result in a reduction of tensile stress, and wn!d not be more limiting. Furthermore, the flaws were assumed to nave high eccentricity ratlos and to be located close to the surface (i.e. higher fluence) which will result in more conservative calculations. "

Page 6 & 10. Labeling has been truncatedinfigure. 1 i

There are stillsome truncatedlabels in thefigure. !

Fixed. '

Page 7. Table ofcontents: "MA TERIALS AND... "should be labeledas Appendix A.

Appendir A and Appendix B are not consistent in Table ofcontents.

1 Fixed.

- Doc a .MED<cclA-64A ATTACHMENT 3,3 PAGE 4 OF- //

Page 12. Why weren't any ofthe thermal transients lookedat such as LOFWP andshutdown cooling?

During intermediate hydrotest condittom, the vessel may not be at isothermal conditions. 1 A LOFWP transient was analyzed for surface welds in the middle to lower vessel regions. An ANSYS finite element analysis was conducted which modeled the vessel wall experiencing a thermal transient due to a 1 OFWP condition. The maximum resulting bending stress due to this transient was found to be 9 ksL This bending stress was then applied in the flaw analysis as an additional weld residual stress; for this particular analysis, the clad residual stress was taken to be zero, since continued normal operation over time will result in a relaxing of the clad residual stress due to creep.

The resulting allowable flaw sizes were found to be more limiting (than the ones determined for hydrotest conditions). Flaw curves have been corrected.

Need detail andjustificationfor assuming hydrotest is worst condition. There can be some thermalstress during some ofthe thermal events which would bepossible based on the Cooper P-Tcurve.

Same asprevious comment. >

See above.

Why is the weldresidual stress perpendicular to the weld assumed to be zero?

Forflaws perpendicular to the weld the report states that the weldresidualstress is not included However, ofaflaw is in the wid itself regardias oforientation, it is subject .~o weld residualstress. Iagree that the stress does decrease significantly, thus reducing the drivingforcefor a longflaw, but that stress is still therefor a shorterflaw. Since theflaw acceptance is based on stress intensityfactor (LEFM) at the ?

deepest point, I believe the stress should be included Resolved with Structurallategrity as per previous response to comments," Letter from W. Lai to A. j Bacha,' Cooper Nuclear Station RPV Flaw Evaluation Handbook Draft Report No. GENE-523-008-0194 (August 1998),' September 10,1998" l

1 Do the eccentricity ratios take into consideration subsurfaceflaws on the ID halfanu OD halfofthe vessel wall?

It is very conservative to use the maximum eccentricity value.

Actually," maximum" eccentricity values were not used. Conservative values were chosen, though, to facilitate calculations. I Page 14. " maximum primary membrane stress"shouldbe " maximum primary local stress".

Corrected.

Page 16. Comment only, no action: Although individualstresses may be elastic, the net stress may be plastic andrequire more sophisticatedevaluation.

N/A Page 17. You may want to state that K* ,cladisfor a longflaw.

I Done.

l Page 27. Firstparagraph, "normaland upset"shouldread " normal, upset, emergency, faulted".

I 1

l a

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' DOC # NOMS-ddb ATTACHMENT 3 e 7>

PAGE G OF IO IM6 Page 43. Provide details ofCode requiredreinspection. Y This was requestedby the customer.

Reference to proper ASME section is provided. This report does not summarize the contents of that ,

section.

{

Provide detailed Code referenceforfaw removal or weld repair.

This was requestedby the customer.

Same as above.

Page 47. Why use minimum wall thickness everywhere? Doesn't the nominalthickness vary at some locations up to 6.375". This may be very conservative.

Isn't there a wid at the transition betueen the 6.375"and 5.375"in the beltline? This does give dgerent stresses and distributions than thatfor a constant thickness shell.

A finite element analysis was performed for a generic case with a wall thickness transition. Results show that stress concentrations occur locally at the " corners" of the wall thickness transition, which are located on the outside surface. Outside surface flaws are not analyzed for the H12 weld, where the thickness transition occurs (any detected outside surface flaws in the H12 weld will require location-specific analysis). For the inside surface, allowable flaws are already governed / limited by IWB-3500. For subsurface flaws, allowable flaw sizes take into account fluence in the beltline region, which is the most limiting condition; subsurface flaws near the outside surface of the vessel wall will l experience lower fluence levels and, unless they "become" outside surface flaws, will not experience l

the local stress concentration due to change in wall thickness. 1 Flaw Diagrams: The mas allowable curves could be higherfor circumferentialfaws.

Was a Cooper specific stress analysis performed to determine the bolt-up stresses using the approp? late bolt loads, dimensions, sealloads, etc. This is unknown since the report does not contain any ofthese details.

A general RPV stress analysis was performed for Fitzpatrick, and results of this analysis were used for bolt-up stresses.

Page 60. Thefaw should be broken into arlal and circumferential components and analyzed both ways.

Agree.

l Appendit B ODfaws were only evaluatedfor weld VFW, state so in Appendix B.

Done.

/ -

Doc c h ED4 A A _ ogg ATTACN e gT - ~L 5 September 10,1998 W (^ _ _0F _ ll l

l Mr. AliBacha Cooper Nuclear Station '

Design Engineering Department

Subject:

Cooper Nuclear Station RPV Flaw Evaluation Handbook Draft Report No. GENE-5234)08-0194 (August 1998) '

i

Dear Mr. Bacha:

This letter is written in response to comments documented in the letter dated August 25,1998, "Tidrd j Party Review of GE Flaw Acceptance Handbook: GE Report GENE-523-008 0194," from Marcos L. I Herrera to Michael J. Friedman. Original comments are shown in Italics. Item by item responses are shown bold.

A final report will be submitted to you via e-mail, followed by a faxed copy of the signed cover sheet. A hardcopy of the report will be sent to Joe Hewett at the following address:

Joe Hewett, GE c/o Cooper Nuclear Station 3.5 Miles off Route 136 i Brownville, Nebraska 68321 ph: 402 825-5018 If you have any questions, please call me at 408-925-4011. Thank you.

Regards, Wilson Lai, GE

. enclosure: response to comments v

ns

/

Doc o M EN9B-o4A ATTACHMENT -- L :s PAGE - 7 _ or _ 11 Executive Summary Page1 Signature blocks " prepared by", " vert)ed by", and " approved by" not signed Final draft will have signatures.

Page 3 There is nojustifcation provided in the report regarding the determination of l limiting events. There is no detailed discussion ofevents other than boltup and hydntest.

Studies were previously pe: formed for a similar plant. Tbc calculations postulated a maximum allowable flaw size based on hydrotest conditions, with an applied stress intensity factor versus allowable stress intensity factor ratio of 1.0 (i.e.

Kg/Kg, =1.0). This same flaw size was used, and conditions of various other limiting events (for example, emergency condition) were applied. The resulting applied stress intensity factor was compared with the allowable stress intensity factor. In each case the resulting ratios were greater than 1.0. It was therefore l shown that boltup or hydrotest events govern, and that resulting fracture toughness j values for these conditions are much lower than those calculated for any other j condition. Reference to this study is included in our design record files.

What is the referencefor the P-T curvefor 16EFPY?

Reference for P-T curve is based on Reference 12 of the report, and is consistent with our conversation with Mike Friedman and All Bacha.

Page 4 Change "allowablefracture toughness" to " allowable stress intensityfactor. "

Done.

Page3 Spell out all words.

Done.

Explain note c (IWB-3500 evaluation is limiting).

The note now makes a reference to Section 3.1.1, which explains IWB-3500 evaluation.

Page 6 & 10 L.abeling has been truncated inpgure.

Corrected.

Page 7 Table ofcontents: "AfA TERIALSAND.. " should be labeled as Appendix A.

Corrected.

Page 8 Disposition time is signo)cantly reduced.

Inst sentence. what does this mean?

Sentence was unclear and superfluous.12 has been <!c!cted.

l 1 I

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we o Aenas.o4s ,

ATTACHMENT _ 3,3 First bullet, were boltup stresses included?

-M- II Boltup stresses were considered, but found to be not limiting.

Page11 Label top scale as " Azimuthal Location. "

Done.

Label " Beltline Region. "

Done.

Page 12 l* paragraph, Boltup stresses? ,

i l

Boltup stresses were considered at the VFW region. This is now stated on page 12.

Why weren 't any ofthe thermal transients looked at such as LOFWP and shutdown cooling?

l Same as "Page 3" response.

Need detail andjustificationfor assuming hydrotest is worst condition. The?e can be some thermal stress during some ofthe therma l events which would be ;

possible based on the Cooper P-Tcurve. l l

Same as "Page 3" response.

Page 14 The weldresidual stress was assumed to be bending. Does this mean tensile on ID and compressive on OD? The residual stress shouldfollow a sinusoidal behavior, tensile on both ID and OD and compressive in the wall center. This may be non-conservativefor ODflaws.

Appendix E of ASME Section XI used linear bending distribution, consistent with calculations in this flaw evaluation report. Furthermore, tensile (Le. positive) stresses wem conservatively applied to both ID and OD weld evaluations. (Note that only VFW considered flaws on the OD weld. Flaws found on any other welds will require location-specific analyses).

Why is the weld residual stress perpendicular to the weld assumed to be zero?

The residual stress along the length of the weld is expected to be tensile in the weld itself but is expected to drop to small compressive magnitude at a small distance in the base metal. On the other hand, the weld residual stress in the transvene direction to the weld persists for a significant distance beyond the weld crown.

This trend has been illustrated in the measured weld residual stress data reported by Ferrill, Juhl and Miller (Measurement of Residual Stresses in a IIcavy l Weidment, Welding Research supplement, November 1966, pp 504s-514s).

l On the basis of the preceding, the residual stresses were not exphtitly included for the postulated flaws transverse tc the weld length.

K,n shouldbe consistent with Kan Corrected.

l l

L L

f/

0000 O ECM % -o O l ATTACHMENT 3*6

! PAGE 9 0F II i Page 16 Addbullet "Vesselweldresidualstress distribution andmagnitude "

l Done.

Do the eccentricity ratios take into consideration subsurfacepaws on the ID half and OD halfofthe vesselwall?

Maximum eccentricity ration were assumed for the purpose of calculating stress correction factors. The specific location of the flaw (ID half or OD half) was not considered.

Page 19 The K ranges may be non-conservative ofthere are some compressive thermal stressespresent (during cooldown on OD).

Fatigue crack gmwth curves are based on Kmin/Kmax ratios (i.e. "R" ratios). The curve associated with the maximum R-ratio is used for the purpose of calculating fatigue crack growth for this report. The K range may increase if Kmin is less than zero. However, ASME Section A-4000 states that if Kmin is equal to or less than zero, then a curve associated with R = 0 should be used. Therefore, the use of a high R-ratio should produce a greater fatigue crack growth value than a low R-ratio curve, for the calculated K ranges. Furthermore, fatigue crack growth were deterrdned to be low, so any differences in K ranges is expected to have minor impact on fatigue crack gmwth.

Page 20 Any basisfor 10 cyclesTeryear?

10 cycles per year is basci on averaging an estimated 400 cycles over a 40 year period, consisting of startups and shutdowns, scrams, and transients.

Page 22 Thefigure shows the clad stress as distributed, but the text states that it was a point load at the cladmidpoint.

In reality the clad stress is distributed through the clad thickness. This stress was, however, reduced to a point load for the purpose of calculating stress intensity.

Page 25 & 63 S/dshould be Sta This should be correct as is.

Page 27 1" paragraph, it is stated that curves are limitingfor all normal and upset events.

What about other events, C & D?

Same as "Page 3" response, Do these curves applyfor both axial and circumferentialpaws?

Curves have been revised to show axial and circumferential flaws.

Do the surfacepaw curves apply to both ID and ODpaws?

Vessel flange weld OD flaw has been included as Figure 3-13. All other detected j OD flaws will require location-specific calculations.

t

'f-DOCO W 5' ~

ATTACHMENT --

II PAGE- I O Of Was crack growth subtractedfmm the U3 limit curves?

Yes, if the allowable flaw size is bounded by the 1/3 limit. If the allowable flaw size does not reach the 1/3 limit, then the 1/3 limit is not reduced for fatigue crack growth, and is shown for reference only.

Where are the curvesfor the dferent eccentricity ratios?

Maximum eccentricity ratios wen assumed for the purpose of calculating stress correction factors (Le. conservative). Therefore, only single sets of can'es are shown for the various flaw orientations and locations.

Last sentence, change to " procedure, pow chart, and worksheet" to reject sequence laid out in handbook.

Done.

Page 34, 35 Jo,37 Correct x andy axis and labels so they don't conpics.

Done.

Page 39 ' Figure 3-10 should read " Cooper Subsurface "V2" Flaw-16EFPY" Done.

Page 42 A dd " emergency" to considered limiting loading.

Done.

Page 43 Provide details ofCode required reinspection.

Reference was made to ASME Section IWB-3132 in the report.

Provide detailed Code referenceforfaw removal or weld repair.

Reference was made to ASME Section IWA-4000 in the report.

Page 47 Why use minimum wall thickness everywhere? Doesn't the nominal thickness vary at some locations up to 6.375". This may be very conservative.

VI weld evaluations have been avised for the 6.375 inch venel wall thickness.

Revised flaw charts am included in the report.

Page 47, 49 -

50 Referencefor ART, CF, RTun. calculations andprovide copyfor review.

References for these values are shown in Ref's. 3,11, and 12 of the flaw evaluation report. Customer has copies of these reports. Calculation of adjusted reference temperatures are based on NRC Reg. Guide 1.99 (i.e. Ref. 9 of flaw evaluation report).

Doco_NED4 'l8 646 ATTAC m e g _ H , :=,

PAGE - 1i . _ op _ t t Page 47 Section A3 needs to be vsworded.

Done.

Page 48 The exponent should be -0.24x, not -0. 024x.

Corrected.

Fluence shouldbe given as E>1MeV.

Done.

Doyou mean the clad inner surface or cladFLASinterface? Areyou taking credit for the attenuation through the clad thickness?

l The vessel inner wetted surface means the " clad inner surface", not the clad /LAS :

interface. Credit is taken for clad thickness.

" fluence,a ,,,,,a" is normalized correctionfactors notfluence level.

This has been clarified in report.

Page 30 RT,,a shouldbe mean RT,,a.

This has been clarified in report. f Page $4 What is the basisfor " conservatively assumed zerofor non-beltline welds"?

Initial RTndt for beltline welds were calculated to be -50 degrees based on material i t

tests. No specific tests were performed for non-beltline weld material Based on historical data (for other plants), initial RTndt for non-beltline weld region can be taken to be no lower than -50 degrees. Therefore, an assumed value of zero would be conservative and appropriate.

Table A-2, V3 should read V3a, V3b, V3c. V4 shouldread V4a, V4b, V4c.

l Done. I 1

Page 60 Per the ASME Code, both the vertical and horizontal components need to be evaluated.

l We agree. Charts now show both axial and circumferential allowable flaws.

Page 62 What is a " Site Corrective Action Program Activity?"

" Site Corrective Action" generically refers to site-specific inspection / reconciliation processes; the customer will incorporate this flaw handbook report into their inspection process documentation. The term " Site Corrective Action Program Activity" has been deleted for clarity.

Page 6? Outside surfaceflaw (top head... Top head was not evaluated in this study. Also on Page 68.

Corrected.

l l

DOC a N EDc 90 - o48 ATTACHMENT - 3-3 Bacfa, All " '

Fr:m: Lal, Wilson C. (NE) [ wilson.lai@ gene.GE.com]

S:nt: Wednesday, November 04,1998 3:07 PM Tc: Mehta, Hardayal S. (NE); Bacha, Ali l l

Cc: Thomas, Kenneth B.; 'Marcos Herrera'

Subject:

RE: CNS Flaw Evaluation Handbook Ali, Section ill of the ASME was used to justify the use of the 1/3 limit for flaw size. The reasoning behind this (i.e. that primary stresses must be less than 1.5Sm) has not changed from the 1965 code.

S ction ill was also used to determine some material properties, which wtre used to calculate input stresses. Since these stresses were conservatively estimated, minor differences in material properties between the 1965 and 1989 codes are negligible.

Dun to the reasons stated above, it is not necessary to modify the fliw handbook to recognize the 1965 code.

Regards, Wilson Lai

> From: Bacha, Ali[SMTP:a1bacha@nppd.com]

> Sent: Wednesday, November 04,199812:20 AM

> To: Lai, Wilson C. (NE); Mehta, Hardayal S. (NE)

> Cc: Thomas, Kenneth B.;'Marcos Herrera'

Subject:

CNS Flaw Evaluation Handbook

> Lai Ken Thomas commented on the use of ASME Code Section Ill,1989 Edition

> as

> noted in the Handbook. Per Ken the CNS RPV is designed to ASME Code,

> Section

> 111,1965 Edition. A later Edition could be used if it is more

> conservative

> (which is uscally the case). How would this affect the Handbook

> r:sults? I

> cm in the pucess of finalizing the calculation and need this

> information

> ASAP. Pleace advise if this is a typo so that I can add a

> clarification in

> my calculation.

> Rtgards

> Ali Bacha

> 402-825-5467 1

( -

H:) CNSPROCS \ FORMS i VOL3 \ 34-7 \ 34-7 3 Ntbraska Public Powst District DESIGN CALCULATIONS SHEET Sheet 1 of

~

NF'JC 98-048 Prepared By:. General Electric (GE) Checked / Reviewed By: Ali Bacha flev, No. O Date.3/1 R 19 98 ' >ate: 11/01 19 98

)

1

{

J l

l I

l

)

l f

i ATTACHMENT 3.4 NPPD Intra-District Memos l

l I

1 l

r l Doc , NE04fg,ot/g ATTACHMENT

.3 , 4 El-1093 xxx PAGE / Of 2 l XEBRASKAPusucPOWERDISTRICT i

l l

l l Date: October 7.1998 Tc: Mike Boyce FOR INTRA-DISTRICT Frum: Spencer Moore

Subject:

Vessel Effective Full Power Years PED 983080 l In accordance with Procedure 3.20 and ENG P97-0009 the following information is provided for the period ending May 31,1998.

Gross Thermal Energy Generated: 328,927,523 MWH Effective Full Power Years (EFPY): 15.77 Where Effective Full Power Years = Gross Thermal Enernv Generated (MWH) 2381*24*365 This satisfies the requirements of Procedure 3.20, step 8.6.2. If you have any further questions contact Spencer Moore, x-5431.

Spencer Moore Engineering Coop cc: Scott McAllister Ole Olson ENG P97-009 File I

N Nebraska Public Power District man En,,o .,oser r

[

ooc o NEDVILoVR ATTACHMENT & W EMcha,Ali PAGE 2- 0F A From: Thomas, Kenneth B.

Sent: Tuesday, October 06,1998 9:35 PM Tc: Bacha,All Cc: Sanjanwala, Narendra M.; Adelung, Perry K.; Martin, John M.; Friedman, Mike J.

Subject:

RE: Executive Summary I can confirm that the pressure test procedure,6. MISC.502, follows the pressure temperature curves in the Technical Requirements manual, and that after the 200 psig inspection, the wssel is heated to the test temperature before it is pressurized to 1005 psig. Operation in the areas to the left of the curws is not permitted. Therefore, the statement in the flaw evaluation handbook is adequate.

-Original Message--

From: Bacha,All Sent: Thursday, October 01,1998 9:12 PM To: Thomas, Kenneth B.

Cc: Sanjanwala, Narendra M.; Adelung, Perry K.; Martin, John M.; Friedman, Mike J.

Subject:

Executive Summary File: letter response 2_. doc >>

Ken i

Please note the comment that GE (Wilson La0 has in the attached letter. The comment is in bold which f reads as follows: "Although the PT curves used for this analysis allow for reactor operatica in areas deemed I more limiting than those used in our calculations (e.g. at 120*F and 610 psi, or at 90*F and 312 psi), these pressure / temperature conditions were not considered, because reactor operation is not expected to reach these zones on the PT curve. The flaw calculations used for this report are based on the expectation that Cooper reactor operation (i.e. power ramp-up and rampdown) will never reach the limiting zones suggested above.

Cooper documented procedures will need to be reviewed, and if they are inconsistent with this assumption, then GE will need to be notified immediately" i Please let me know if there are any concems with this comment. .

Regards Ali Bacha i

i l

l l

i i

i l

l 1

H:) CNSPROC8 \ FORMS WOL3 % 3 4-7 n 3+7.3 NIbriska Public Powor District DESIGN CALCULATIONS SHEET sheet nor Z NEDC 98&B Prepared By: General Electric (GE)/NPPD Checked / Reviewed By: Ali Bacha Rev.No. 0 Date: 10/15 19 98 Date: 11/01 19 98 e

ATTACHMENT 3.5 Analysis of UT Indications in the ,

Reactor Pressure Vessel (RPV) Shell Welds i

l l

H:) CNSPROCS \ FORMS WOL3 % 347 \ 3+7.3 N:br:ska Public PowIr District DESIGN C LCULATIONS SHEET sheet .!_ or _8 e i NEDC 98-o48 Prepared By: Ali Bacha -

Checked / Reviewed By: John Marti#M Rev. No. o Date: 11/o9 19 98 Date: llatl 19 98 Analysis of UT Indications in the Reactor Pressure Vessel (RPV) Shell Welds 00c e annM2-04'A ,

1.

Purpose:

ATTACHMW M y / gp g i PIR 3-50914 (Condition Report CR98-0882) documented the condition that ultrasonic testing (UT) of the Reactor Pressure Vessel (RPV) welds VLA-BA-3 and VLC-BB-2 indicated that some flaws exceeded the IWB-3500 flaw allowable size requirements of ASME Code Section XI The purpose of this evaluation is to evaluate the unacceptable flaws in the welds to the requirements ofIWB-3600 of ASME Code,Section XI and to provide closure to the PIR. The unacceptable indications documented in GE's data report numbers RPV-03 and RPV-10 are the subject of this evaluation.

2. Affected System (s): Reactor Pressure Vessel (Nuclear Boiler)

3. Affected Documents: None

4. Requirements / Design Inputs:

The reactor vessel is classified as an essential system. The vessel provides a volume in which the core can be submerged in coolant allowing power operation of the fuel.

The reactor vessel shell and its associated welds are designed to class I loads per the requirements of CNS USAR.

5.

References:

5.1 PIR 3-509)4 5.2 GE's Flaw Evaluation Handbook Report Number GENE-523-008-0194 Revision 1 5.3 UT ISI Inspection Report Numbers RPV-03 and RPV-10 5.4 ISI Summary Inspection Report for Refueling Outage 18 (RFOl8)

6. Attachments:

See attachments 3.6 and 3.7 for UT Reports and corresponding flaw evaluation worksheets.

7. Method:

{ The method used in this evaluation is based on the guidelines of the Flaw Evaluation Handbook (attachment 3.2 ofNEDC98-048) AND are consistent with the Section XI, ASME Code,1989 Edition.

H:% CNSPROCS % FORMS WOL3134 7 4 34 7,3 N1br ska Public Powu District DESIGN CALC LATIONS SHEET sheet .g. .e .T NEDC 98-o48 Prepared By: Ali Bacha Checked /Reviewad By: John Marti[

Rev. No. o Date: 11/o9 19 98 Date: ll ll 19 98 !

noc e 4GIL AB - 04B ATTACHMENT MJ

8. Description of what is being evaluated: PAGE 2. OF 8 l

The Inservice Inspection (ISI) ultrasonic (UT) examination of CNS's RPV shell welds was perfonned by General Electric (GE) during refueling outage 18 (RFOl 8).

The RPV shell welds were examined to satisfy the American Society of Mechanical Engineers (ASME),Section XI requirements. The ASME Section XI required l

' examination volume of the RPV welds was examined with the GERIS 2000 Invessel l System from the RPV inside surface. The GERIS 2000 is configured so that the acceptance standards of ASME Code XI, Table IWB-3510-1 flaw size allowables are built-in for inunediate analysis of the UT readings for flaw indications. The indications that exceeded the IWB-3500 flaw size allowables are required to be analyzed per IWB-3600 (1989) Edition. This evaluationjustifies acceptability of the indications to the requirements ofIWB-3600.

9. Detailed Evaluation:

The UT examination of the RPV shell welds detected a number of recordable indications. A total of thirteen (13) flaw indications were recorded in Report Number !

RPV-03 and a total of forty nine (49) flaw indications were recorded in Report Number RPV-10. All of the indications are close to the mid-plane and appeared to be fabrication related. As indicated on the GE's "GERIS 2000 Invessel Inspection Summary Sheets" the flaws are attributed to slag remaining from insufficient back gouging during fabrication. All indications that required evaluation to IWB-3600 are characterized as subsurface in nature in accordance with IWA-3320 and Figure IWA-3320-1 and were combined in accordance with IWA-3330. The distribution of the recordable indications is shown in the relevant attached exam sheets a:.d drawings.

9.1 Description of the RPV-03 Report Flaw Indications:

Report Number RPV-03 (attachment 3.6) included the summary of the l GERIS 2000 Invessel weld inspection results for weld number VLA-BA-3 i (reference attached dwg. CPR-0001). The inspeded weld is a vertical weld ,

located approximately between elevations 110.0" (elev. 926') and 253.0" i (938') (0.00" (917') being the invert level of the vessel bottom head) along ,

Azimuth 258*. A total of thirteen (13) flaw inJications were recorded in report number RPV-03. The indications ranged in length (L) from 0.50" to 1.50". The indications had through-wall (2a) dimensions that ranged from 0.184" to 0.55". The indications had varied elevation locations. The indications were subsurface with a minimum "S" distance (ligament) of approximately 2.90". Of these thirteen (13) flaw indications four (4) i exceeded Table IWB-3510-1 allowable sizes. The four indications are characterized as subsurface indications, identified as 3-006,3-007,3-008, and 3-009, and are located between an elevation band of 187.10" (932.6') and

H:1CNSPROC8% FORMS 1 Vola)34 7%3-4-7 3 N;br::sks Public P4 wit District DESIGN CALCULATIONS SHEET sheet 5_ or f NEDC 98-o48 ' Prepared By: Ali Bache Checked / Reviewed By: John Martir h i ' T l

R.v. No. o Date: 11/o9 19 98 Date: _ ll ll 19 98

(

9.1 Description of the RPV-03 Report Flaw Indications: (continue) 193.60"(933'). Indications 3-006/3-007 and 3-008/3-009 were combined per the proxituity rules of the requirements ofIWA-3330 of the ASME Section XI and justified as unacceptable for not meeting the IWB-3500 allowable flaw size requirements. This is documented in the attached sheets of the 1 examination report (attachment 3.5).

9.2 Description of the RPV-10 Report Flaw Indications:

~ Report number RPV 10 (attachment 3.'/) included the summary of the GERIS 2000 Invessel weld inspection results for weld number VLC-BB-2 (reference attached dwg. CPR-0001). The inspected weld is a vertical weld located approximately between elevations 403.0" (elev. 950') and 553.0" (963')

(0.00" (917') being the invert level of the vessel bottom head) along Azimuth j 120*. A total of forty nine (49) flaw indications were recorded in report number RPV-10. The indications ranged in length (L) from 0.25" to 2.50" (with two indications of the forty nine indications recorded at 3.0" and 5.25").

The indications had through-wall (2a) dimensions that ranged from 0.141" to 0.311". The indications had varied elevation locations. The indications were subsurface with a minimum "S" distance (ligament) of approximately l 1.00". Of these forty nine (49) flaw indications seven (7) exceeded Table i IWB-3510-1 allowable sizes. The seven indications are characterized as i subsurface indications and are identified as indications10-002,10-003,10-004,10-010,10-011,(all five are located at the bottom portion of the weld),10-048,10-049 (both located at the top portion of the weld). Indications 10-002/10-003/10-004,10-010/10-011, and 10-048/10-049 were combined per the proximity rules of the requirements ofIWA-3330 of the ASME Section XI and justified as unacceptable for not meeting the IWB-3500 allowable flaw size requirements. This is documented in the attached sheets of the examination report (attachment 3.7).

9.3 Flaw Evaluation Worksheets The flaw indications noted in 9.1 and 9.2 above are further evaluated per the requirements ofIWB-3600, ASME Code- Section XI. The " Flaw Evaluation

' Handbook", documented in attachment 3.1 of this calculation, provided the necessary worksheets to document the IWB-3600 evahiation and to compare the results with the flaw size allowable charts provided in Figures 3-1 to 3-16 of the calculation. The flaw size allowable charts shown in Figures 3-1 through 3-16 consider the flaw allowable sizes calculated using end oflife (EOL) fatigue crack growth values. These worksheets are included in attachments 3.6 and 3.7 for welds VLA-BA-3 and VLC-BB-2 respectively.

ooc , MErr % 048 ATTACHMENT 39 nnr 3 ne G

I~

l l

H: % CNSPROCS i FORMS \ VoLS % 3 4-7 % 3-4 7,3 Nibriska Public Pcwor District DESIGN CALCULATIONS SHEET sheet 4' or S~

NEDC 98-o48 Prepared By: Ali Pacha Checked / Reviewed By: John Mar' Rev. No. o Date: 11/o9 19 98 Date: 11. ICP 19 98 v

10 Summary review of the inspection results l

By review of the "GERIS 2000 Invessel Summary Sheets" for report numbers RPV-03 and RPV-10 results for the inspected welds and based on the above evaluation, the following is noted:'

j 10.1 Report number RPV-03 (weld VLA-BA-3) l l a- Approximately 69% of the 13 reported indications are acceptable per ASME i

Code Section XI-IWB-3500 requirements. The unacceptable 31% of the 13 reported indications arejustified to meet the requirements ofIWB-3600.

b- Previous data was reviewed prior to the completion of GE's summary sheet.

Report number R-223 (1995) recorded indications evaluated as being acceptable correlating to the unacceptable indications. The change in evaluations is attributed to changes in the examination surface and equipment.

l c- The unacceptable flaw indications were plotted to the region of the weld root l and are attributed to slag remaining from insufficient back gouging during fabrication.

l 10.2 Report number RPV-10 (weld VLC-BB-2) l a- Approximately 86% of the 49 reported indications are acceptable per ASME l Code Section XI-IWB-3500 requirements. The unacceptable 14% of the 49 reported indications are justified to meet the requirements ofIWB-3600.

! b- Previous data was reviewed prior to the completion of GE's summary sheet.

l The areas containing the unacceptable indications were not accessible during I

the previous examinations.

I c- The unacceptable flaw indications were plotted to the region of the weld root l and are attributed to slag remaining from insufficient back gouging during

! fabrication.

1

\

11.

Conclusion:

Flaw indications 3-006, 3-007, 3-008, and 3-009 (ref. VLA-BA-3) AND flaw indications10-002,10-003,10-004,10-010,10-011,10-048, and 10-049 (ref. VLC-BB-2) are found to be acceptable per the requirements of ASME Code, Section XI-IWB-3600. As discussed above,it is concluded that the above flaws are not service i induced flaws and the flaws are attributed to slag remaining from insufficient back gouging during fabrication.

Doc o NGCE00"OA0 ATTACHMENT E9 PAGE 4 0F X L. j

I y H:t CNSPP'AS i PORMS WOL3 \ 34 7) 34 7,3 NIbr:ska Public Powor District l DESIGN CALCULATIONS SHEET sheet _S or _6 NEDC 98-o48 Prepared By: Ali Bacha Checked / Reviewed By: John MarfW f

Rev No. o Date: 11/o9 19 98 Date: ll. l(2 19 98~

k; ooc,MER.M.046

11.

Conclusion:

(continue) p t &

ASME Code Section XI, IWB-3610 states that flaws exceeding the allowables l defined in IWB-3500 may be evaluated by analytical procedures to calculate growth until the next inspection or the end of service lifetime of the component. Based on l the analysis presented above, the allowable flaw sizes used to justify the acceptability of the flaw indications to IWB-3600 were calculated using end oflife (EOL) fatigue i crack growth values. Therefore, based on the results of the analysis, continued I operation 'as is' is acceptable for the remainder of plant life.

l i

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i H: \ CNSPCOCS \ FORMS iVOL3 % 34-7 \ 347,3 Nibr ska Public Powar District DESIGN CALCULATIONS SHEET sne.t or _-

NEDC 98 & 8 Prepared By: General Electric (GE) !N f[9 Checked / Reviewed By: Ali Bacha i Rev No. O Date: 10/16 19 98 Date: 11/01 19 98 ATTACMENT 3.6 UT Report No. RPV-03 and Corresponding Flaw Evaluation Worksheets 1

i l-

i GE Nuclear Energy GENE-523-008-0193 Rev.1 Doc o 0606 %~W '

AT TACHMENT - 3*b PAGE / OF 6 COOPER NUCLEAR STATION FLAW EVALUATION WORKSHEET '

4 l FlawID: VLA-BA-3. Indications 3-006/3-007_y j 1

1) Determine Region and Orientation of Flaw. The weld region should be identified by the nearest weld. The orientation is either (A]xial or [C]ircumferential. If the I flaw is at a junction between two welds, the region with the more limiting acceptance criteria should be conservatively used.

l Region: V1 Orientation: Axial i

I

2) Sketch Flaw Geometry. Use the attached flaw sketch to draw the flaw. l l

3) Classify Flaw. Combine flaws in close proximity to other flaws and to the surface per the proximity rule ofIWA-3300,Section XI of the ASME Code. Classify flaw as either:

Inside Surface Outside Surface Subsurface X

4) Determine Vessel Wall Geometry. If the flaw is classified as subsurface or outside surface, input 0 for clad thickness, else enter the analysis value for clad thickness as listed in Table A-1 of Appendix A for the specified weld region.

Cladding Thickness, tclad = 0 (in)

Low Alloy Steel Thickness, tLAS = _7.00 _ (in)

Total thickness, t=tclad + tLAS = _ 7.00 (in)

A RJerce weld no- V 1.c we \d 1-2 Mc[r,p,,,

,-2 *R,Hg dq "'

ge. io J w.%suacmums39 J

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r' .~.

GENuclear EnerU GENE-523-008-0194 Rev.)

DRF # 137-0010-7 (Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID:_ VLA-BA-3.

Indications 3-006/3-007 Doc e 4EM984 ATTAcuussi A.6 PAGE 2- 0F /Y

5) Size Flaw. Calculate flaw depth, including any portion of the flaw extending into the cladding.

l Surface Flaws: Subsurface Flaws:

Flaw Depth, a = _N/A (in) Flaw Depth,2a = _0.55___ (in)

Flaw Length, L = N/A (in) HalfDepth, a = 0.275 (in)

Flaw Length, L = 2.0 (in) l

6) Calculate Aspect Ratio of Flaw. ;

I Flaw Aspect Ratio, a/L = __0.138

7) IWB-3500 Flaw Evaluation. For the given a/L aspect ratio, determine the I allowable flaw depth, a (surface) and 2a (subsurface), in accordance with IWB-3510 of the Code and record the value below. If the flaw depth recorded in step 5 is below the allowable value, check the box " Acceptable per IWB-3500" below. Otherwise, check the box " Unacceptable per IWB-3500" and continue to step 8.

Inside Surface Flaw:

IWB-3500 Allowable Depth = a = _N/A_ (in)

Outside Surface Flaw (vessel flange regions only):

IWB-3500 Allowable Depth = a = _N/A (in) i Subsurface Flaw:

l IWB-3500 Allowable Depth = . a = _. 0.392_ (in) ( o . M O%'U l

67

i GENuclear Energy GENE-523-008-0194 Rev.1 DRF M i37-0010-7 (Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID:_ VLA-BA-3.

Indications 3-006/3-007__

ooc o NGLE98.-o42 ATTACHMENT - A6 ACCEPTABILITY: " "

Acceptable per IWB-3500. ;

X Unacceptable perIWB-3500.

)

8) IWB-3600 Flaw Evaluation. Record the appropriate flaw acceptance diagram Figure # from Section 3.0. Record the allowable flaw depth, a or 2a, from the

)

appropriate curve for the specified orientation. If the flr.w depth recorded in step 5 is below the allowable value, check the box " Acceptable per IWB-3600" below. Otherwise, check the box " Unacceptable per IWB-3600", and proceed to step 9. j NOTE: Flaw specific analysis would be required if outside surface flaws were found in any region below the vessel flange.

Figure # 3-12_ .

Inside Surface Flaw:

IWB-3600 Allowable Depth = a = N/A___ (in)

Outside Surface Flaw (vessel flange regions only):

IWB-3600 Allowable Depth = a = _____N/A (in)

Subsurface Flaw:

IWB-3600 Allowable Depth = 2a = 1.05 (in) > 0.55 (in)

ACCEPTABILITY:

_X_ Acceptable per IWB-3600. (for 16 EFPY)

Unacceptable per IWB-3600.

l 1

l

. 68 I i

GENuciaar Energy GENE-323-008-0194 Rev.1 DRF H 137-0010-7 l

l i

(Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID: VLA-BA-3. l Indications 3-006/3-007_ oocaM N 48-048 ATTACHMENT d'G PAGE 4 0F . IO

9) From figure identified above, record the 1/3 wall thickness limit below. If flaw l depth is below 1/3 limit, flaw removal is acceptable. Otherwise, weld repair is necessary.

1/3 Limit = N/A (in)

From step 5 above:

Flaw depth = a = N/A__(surface) 2a > N/A (subsurface) t Flaw excavation depth < 1/3 Limit: Flaw removal acceptable (No weld repair)

Flaw excavation depth > 1/3 Limit: Weld repair required 1 \

l l

l r

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- 69 L

c GE Nuclear Enargy GENE-323-008-0194 Rev.1 DRF H 137-0010-7 COOPER NUCLEAR STATION FLAW EVALUATION WORKSHEET 4 4 000 # U N T-00 FlawID: VLA-BA-3. Indications 3-008/3-009 ATTACHMENT 3*b

/6 f

PAGE 8 0F

1) Determine Region and Orientation of Flaw. The weld region should be identified by the nearest weld. The orientation is either [A]xial or (C]ircumferential. If the l flaw is at a junction between two welds, the re~ ion with the more limiting acceptance criteria should be conservatively used.

l l

Region: V1 Orientation: Axial l 2) Sketch Flaw Geometry. Use the attached flaw sketch to draw the flaw.

3) Classify Flaw. ' Combine flaws in close proximity to other flaws and to the surface !

per the proximity rule ofIWA 3300,Section XI of the ASME Code. Classify flaw as either:

Inside Surface Outside Surface Subsurface _X 1

l 4) Determine Vessel Wall Geometry. If the flaw is classified as subsurface or outside surface, input 0 for clad thickness, else enter the analysis value for clad j

thickness as listed in Table A-1 of Appendix A for the specified weld region. )

Cladding Thickness, tc lad = 0 (in) !

Low Alloy Steel Thickness, tLAS =

7.00_____ (in)

Total thickness, t=tc lad + tLAS =

7.00 (in) j A Refe.rtme. *td na. ~41.cWid zuc/A' pre l-1 " f?ot tout drao h ' to 03 %<.- lb4 bete ( AnoAmcJ L i )

4 57,c.fere e D g . 64..

cpg-tott 64 I4. dak shut 2 3-o08/M ;

. 66 o

i -1 1

GENuclear Energy GENE-523-008-Of 94 Rev.1 DRF M 137-0010-7 (Cooper Nuclear Station Flaw Evaluation Wowsheet cont'd) Flaw ID:_ :, VLA-BA-3.

Indications 3-008/3-009 _ p ATTACHMENT 3.O PAGE- 0 0F /5

5) Size Flaw. Calculate flaw depth, including any portion of the flaw extending into the cladding.

Surface Flaws: Subsurface Flaws:

Flaw Depth, a = N/A (in) Flaw Depth,2a = 0.495__ (in)

Flaw Length, L = N/A (in) Half Depth, a = 0.248 (in)

Flaw Length, L = _1.75 (in)

6) Calculate Asoect Ratio of Flaw.

Flaw Aspect Ratio, a/L = __0.141 7). IWB-3500 Flaw Evaluation. For the given a/L aspect ratio, determine the allowable flaw depth, a (surface) and 2a (subsurface), in accordance with IWB-S510 of the Code and record the value below. If the flaw depth recorded in step 5 is below the allowable value, check the box " Acceptable per IWB-3500" below. Otherwise, check the box " Unacceptable per IWB-3500" and continue to step 8.

Inside Surface Flaw:

IWB-3500 Allowable Depth = a = __N/A_ _ (in)

Outside Surface Flaw (vessel flange regions on!y):

IWB-3500 Allowable Depth = a = _N/A (in) i Subsurface Flaw:

IWB-3500 Allowable Depth = 2a = 0.396_(in) < 0.495 (in) ]

67

l GENuclear Energy GENE-523-008-0194 Rev.1 DRF # 137-0010-7 l

(Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID:_ :_ VLA-BA-3.

Indications 3-008/3-009 ~

ooca M E N.9e-o49 ATTACHMENT - 3

Cs PAGE ~7 0F- /8 ACCEPTABILITY:

Acceptable per IWB-3500.

X Unacceptable perIWB-3500.

8) IWB-3600 Flaw Evaluation Record the appropriate flaw acceptance diagram Figure # from Section 3.0. Record the allowable flaw depth, a or 2a, from the appropriate curve for the specified orientation. If the flaw depth recorded in I step 5 is below the allowable value, check the box " Acceptable per IWB-2500" below. Otherwise, check the box " Unacceptable per IWB-3600", and proceed to

)

step 9.

NOTE: Flaw specific analysis would be required if outside surface flaws were found in any region below the vessel flange.

Figure # 3-12 Inside Surface Flaw:

IWB-3600 Allowable Depth = a = N/A (in)

Outside Surface Flaw (vessel flange regions only):

IWB-3600 Allowable Depth = a = N/A (in)

Subsurface Flaw:

IWB-3600 Allowable Depth = 2a = 1.05 (in) > 0.495 (in) l ACCEPTABILITY:

X_ Acceptable per IWB-3600. (for 16 EFPY)

Unacceptable per IWB-3600.

i l

f 68

I.

i GENuclear Energy GENE-523-008-0194 Rev.1 DRF # 137-0010-7 (Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID: _ :V _ LA-B A-3.

Indications 3-008/3-009 -

occ* A E N 99 046 I ATTACHMENT 3- 0 PAGE h DF /I

9) From figure identified above, record the 1/3 wall thickness limit below. If flaw depth is below 1/3 limit, flaw removal is acceptable. Otherwise, weld repair is necessary.

1/3 Limit = N/A (in)

From step 5 above: i Flaw depth = a = N/A__ (surface) 2a = N/A_ (subsurface)

Flaw excavation depth < 1/3 Limit: Flaw removal acceptable

{

(No weld repair)

Flaw excavation depth > 1/3 Limit: Weld repair required 69

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DRF # 137-0010-7 I

DOC , nw B MB i ATTACHMENT 3=0 PAGE /d 0F M Azimuthal Location 0

30 60 90 120 150 180 210 240 270 300 I 330 I l g ;

I VFW Weld l

l V4a Weld V4b Weld V4r Weld

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. ...... .......................(./

........................ ... .... . . . .. .... .11 1. 2..W.. .e.ld.... ......

Via Weld Vib Weld Vic Weld

[2-233A [2 233B [2 233C j

i FIGURE l-2 Cooper Nuclear Station Partial RPV " Rollout" Drawing I

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. 9 GE Nuclear Energy '

GERIS 2000 INVESSEL

SUMMARY

SHEET REPORT NO.:

i PROJECT: COOPER UNIT 1_ RE18 PROCEDURE: GF UT-7n0V2 REV._bl/A. FRR: 1B8P3-001

__N/A

_ N/A SYSTEM: _BEACIORPBESSURE_ VESSEL PROCEDURE: GF-UT-701V2 REV._N/A FRR: __ N/A N/A WELD NO.: ._VIA-BA-3 N/A PROCEDURE: t1T-CNS-300V3 REV.O FRR: 1G08L-001 CONFIGURATION: _LONGIIUDINAL N/A N/A DATA SHEET NO.(S): Examination Data Sheets 341 thru 3-07, Indication Data Sheets 3-001 thru 3-015 Indication Screcn Prints 3-001 thnJ 3-015, Indication Plots, Indication Evaluation Sheets. Exam Patch Location Map v - Coverage Data and GERIS 2000 invessel Setup Records Manu' !bration Sheets RC-034, RC-035, RC-036. Manual Data Sheets RD-034, RD-035, RD436 I I

Weld VLA-BA-3 was examined utilizing O' longitudinal wave,45' shear wave,60* shear wave and 70* refracted longitudinal (RL) I wave techniques. I The ASME Section XI required examination volume was examined with the GERIS 2000 invessel System from the RPV inside surface.

~

l The examination was limited due to the jet pump diffuser to vessel wall clearances, shroud support gussets, and manipulator i

lower scan limit. '

The automated examination coverage was calculated to be 74.9%. I Areas limited to the GERIS 2000 were examined rnanually utilizing O' longitudinal wave 45* shear wave, and 60* shear wave techniques from the RPV outside surface. Access to the OD surface was from the RPV bottom head / vessel skirt region. ,

The manual examination was limited by the insulation support ring and by bioshield access.

gggg,,MQ The manual examination coverage was calculated to be 5.7% (5.7% applied). ATTACHMENT 3*b PAGE Il 0F I The total examination coverage was calculated to be 80.6%.

Thirteen (13) flaw indications were recorded wita the GERIS 2000 and evaluated as being acceptable to the requirements of IWB-3500-1.

Indications 3-006 and 3-007 were combined in accordance with IWA-3330 and evaluated as being unacceptable. fr# /

Indications 3-008 and %009 were combined in accordance with (WA-3330 and evaluated as being unacceptable.

The unacceptable flaw indications were plotted to the region of the weld root and are attributed to slag remaining from insufficient back gouging during fabrication.

One (1) geometric indication was recorded from the outside surface.

One (1) nonrelevant indication due to plate segregates was recorded.

Previous data was reviewed prior to the completion of this summary. Report R-223 (1995) recorded indications evaluated as being acceptable merelating to the unacceptable indications. The change in evaluations is atributed to changes in the examination surface and equipment.

Ths examination results are unacceptable per the requirements of ASME Section XI,1989 Edition No Addenda.

____ __.__.. _ _. d03 Y_ - _ . _ _ _ __..

_ . g _. .. M 48 . ._

SUMMARY

DY: LEVEL oATE NPPD NoE REVIEW 8Y: oATE GER

_ E _ l hf/f/ Pact j_ or &

. LEVEL DATE H.S.BJ & L Co. REVIEW 8Y: oATE e su. arw Report RPV-03 Page I of 93

i DOC 0 N F%90--d4A ATTACHMENT - & Co l PAGE_ / 2 0F /I GERIS 2000 Indication GE Nuclear Energy gygjyggjon gggy ggggg Project : Cooper Nuclear Power Station Exam Data Sheet: 3-05 WoldID: VLA-BA-3 Ind. Data Sheet: 3-00613-007 Patch ID : V1C-02A Indication : 6->7 l

Flaw ThruwallDimension = 0.55 *T* nominal = 6.68 Flaw Length ~1* = 2.00 *T* measured = 7.00 Surface Separation "S* = 3.00 Clad *T* nominal = 0.31 ASME Section XI,1989 Edition, No Addenda TABLE IWB 3510-1 for 4" and Greater l

l l

all Surface % Subsurface % Surface % Subsurface % j 0.00 1.9 2.0 - - 1 0.05 2.0 2.2 - -

l !

0.10 2.2 2.5 2.43 2.80 Y 0.15 2.5 2.9 - -

0.20 2.8 3.3 - -

0.25 ' .3 3.8 - -

0.30 3.8 4.4 - -

0.35 4.4 5.1 - -

0.40 5.0 5.8 - -

0.45 5.1 6.7 - -

0.50 5.2 7.6 - -

Allowed Mowed i 2.43 2.80 a= 0.275 all value = 0.138 Y= 1.000 Flaw is Subsurface Allowed alt = 2.80%

alt = 3.93%

l l Flaw is unacceptable by Table IWB-3510-1.

1 Comments Flaw is axial.

Combined flaws 3-006 and 3-007 per IWA-3330.

Analyst: Reviewed By: N .

O Levelgg Date: f- Level: ' TIT- Date: InI?R/4Tr Report RPV-03 Page 50 of c3

f 00C O $ G 0 0 08-0 40 ATTACHMENT _A G

\ m n Ous l' GERIS 2000 Indication GE Nuclear Energy -

Evaluation Data Sheet Project : Cooper Nuclear Power Station Exam Data Sheet: 3-05 WeldID: VLA-BA-3 Ind. Data Sheet: 3-00813-009 Patch ID : V1C-02A Indication : 8->9 Flaw ThruwallDimension = 0.495 "T* nominal = 6.88 Flaw Length "I* = 1.75 "T" measured = 7.00 l Surface Separation "S* = 2.90 Clad "T" nominal = 0.31 \

ASME Section XI,1989 Edition, No Addenda i l

TABLE IWB-3510-1 for 4" and Greater '

all Surface % Subsurface % Surface % Subsurface %

0.00 1.9 2.0 - -

0.05 2.0 2.2 - -

0.10 2.2 2.5 2.45 2.83 Y 0.15 2.5 2.9 - -

0.20 2.8 3.3 - -

0.25 3.3 3.8 - -

l 0.30 3.8 4.4 - - l 0.35 4.4 5.1 - -

0.40 5.0 5.8 - -

0.45 5.1 6.7 - -

0.50 5.2 7.6 - -

Allowed Allowed I 2.45 2.83 a= 0.248 all value = 0.141 Y= 1.000 Flaw is Subsurface Allowed alt = 2.83%

a/t = 3.54 %

Flaw is unacceptable by Table IWB-3510-1.

Comments Flawis axial.

Combined flaws 3-008 and 3-009 per IWA-3330.

Analyst. Reviewed By: ./

Level: Date: M 7//f Level: Date: 1o bs'/4 Sr Report RPV-0) Page $1 of 93

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r H: \ CNSPROCS \ FORMS \ VOL3 % 34-7 ) 34 7,3 N:brtsks Public Powtr District DESIGN CALCULATIONS SHEET sneet n or _

NEDC 98-048 Prepared By:. General Electric (GE)/M f9 Checked / Reviewed By: Ali Bache Rev.No. O Date: 10/15 19 98 W

Date: 11/01 19 98 4

ATTACHMENT 3.7 UT Report No. RPV-10 and Corresponding Flaw Evaluation Worksheets l

1 I

l l

i

GENuclear Energy GENE-323-008-0194 Rev.!

DRF # 137-0010-7 DOC # M d b- O M ATTACHMENT __,3 * ~I PAGE I 0F M COOPER NUCLEAR STATION FLAW EVALUATION WORKSHEET l Flaw ID: VLC-BB-2 * . Indications 10-048/10-049 *

1) Determine Region and Orientation of Flaw. The weld region should be identified I by the nearest weld. The orientation is either [A]xial or [C]ircumferential. If the flaw is at a junction between two welds, the region with the more limiting l acceptance criteria should be conservatively used.

i Region: V3 l Orientation: Axial

2) Sketch Flaw Geometry. IJse the attached flaw sketch to draw the flaw.

i

3) Classify Flaw. Combine flaws in close proximity to other flaws and to the surface per the proximity rule ofIWA-3300,Section XI of the ASME Code. Classify flaw as either:

Inside Surface Outside Surface Subsurface X

4) Determine Vessel Wall Geometrl if the flaw is classified as subsurface or outside surface, input 0 for clad thickness, else enter the analysis value for clad thickness as listed in Table A-1 of Appendix A for the specified weld region.

Cladding Thickness, tclad = 0 (in)

Low Alloy Steel Thickness, tLAS = _5.875 (in)

Total thickness, t=tclad + tLAS = _5.875 __ (in)

Reference weld no. V3b ,1234B/ Figure 12

Rollout drawing", page 10 of the ((andbook (Attachment 3,1) .

Reference Dwg. no. CPR 101I and exam data sheet 10-04 of attached weld examination sheets.

. 66

I' GE Nuclear Energy GENE-523-008-0194 Rev.1 DRF # 137-0010-7 (Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID: VLC-BB-

2. Indications 10-048/10-049 .

9,, gg ATIAChMENT 37 l PAGE % Of N

5) Size Flaw. Calculate flaw depth, including any portion of the flaw extending into l the cladding. l Surface Flaws: Subsurface Flaws:

l Flaw Depth, a = N/A (in) Flaw Depth,2a = 0.4 (in)

Flaw Length, L = _ N/A (in) Half Depth, a = 0.2 _ (in)

Flaw Length, L = 1.25 (in)

6) Calculate Aspect Ratio of Flaw.

I Flaw Aspect Ratio, a/L = __0.16 i

l

7) IWB-3500 Flaw Evaluation. For the given a/L aspect ratio, determine the allowable flaw depth, a (surface) and 2a (subsurface), in accordance with IWB-3510 of the Code and record the value below. If the flaw depth recorded in l

step 5 is below the allowable value, check the box " Acceptable per IWB-3500" l

below. Otherwise, check the box " Unacceptable per IWB-3500" and continue to step 8.

I Inside Surface Flaw:

IWB-3500 Allowable Depth = a = N/A (in)

)

Outside Surface Flaw (vessel flange regions only) ,

IWB-3500 Allowable Depth = a = !

N/A__ (in) i l

Subsurface Flaw: ,

IWB-3500 Allowable Depth = 2a = 0.35_ (in) < 0.4 (in) l 67 L

L

I GE Nuclear Energy GENE-523-008-0194 Rev.1 DRF # 137-0010-7 l

(Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID: VLC-BB-2.

Indications 10-048/10-049 ooc , Q Eodd- 048 _

~

ATTACHMENT 3*7 PAGE- 3 0F M ACCEPTABILITY:

Acceptable per IWB-3500.

.){._ Unacceptable per IWB-3500.

l 8) IWB-3600 Flaw Evaluation. Record the appropriate flaw acceptance diagram Figure # from Section 3.0. Record the allowable flaw depth, a or 2a, from the appropriate curve for the specified orientation. If the flaw depth recorded in step 5 is below the allowable value, check the box " Acceptable per IWB-3600" below. Otherwise, check the box " Unacceptable per IWB-3600", and proceed to step 9.

l NOTE: Flaw specific analysis would be required if outside surface flaws were found in any region below the vessel flange.

Figure # 3-8 OWB-3600 Axial Flaw Eval. Curve) i Inside Surface Flaw:

IWB-3600 Allowable Depth = a = N/A _ (in)

Outside Surface Flaw (vessel flange regions only):

IWB-3600 Allowable Depth = a = _N/A (in)

Subsurface Flaw:

IWB-3600 Allowable Depth = 2a = 1.75 (in) > 0.4 (in)

! ACCEPTABILITY:

X Acceptable per IWB-3600. (for 16 EFPY)

Unacceptable per IWB-3600.

I 68 i'

g-GE Nuclear Energy GENE-523 008-0194Ra.!

DRF # 137-00.'O-7 l

(Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID: VLC-BB-2.

Indications 10-048/10-049 coc , - AIEpc 9d-o46 AT TACHMENT - 3.7 PAGE- 4 _ or M

9) From figure identified above, record the 1/3 wall thickness limit below. If flaw depth is below 1/3 limit, flaw removal is acceptable. Otherwise, weld repair is necessary.

1/3 Limit = N/A (in)

From step 5 above:

Flaw depth = a = N/A _ (surface) 2a = ___ N/A _ (subsurface)

Flaw excavation depth < 1/3 Limit: Flaw removal acceptable (No weld repair)

Flaw excavation depth > 1/3 Limit: Weld repair required I

. 69 j

p-I

)

l GENuclear Energy GENE-523-008-0194 Rev.1

( DRF # 137-0010-7 \

00C o SHl%$8-d$8 k ATTACHMENT -- 3. ~7 FAGE 8 ,0Fi SO COOPER NUCLEAR STATION FLAW EVALUATION WORKSHEET Flaw ID: VLC-BB-2 e . Indications 10-002/10-003/10-004 *

1) Determine Region and Orientation of Flaw. The weld region should be identified l

by the nearest weld. The orientation is either [A]xial or [C]ircumferential. If the flaw is at ajunction between two welds, the region with the more limiting j

- acceptance criteria should be conservatively used.

Region: V3 Orientation: Axial l

2) Sketch Flaw Geometry. Use the attached flaw sketch to draw the flaw. ~

3) . Classify Flaw, Combine flaws in close proximity to other flaws and to the surface l per the proximity rule ofIWA-3300,Section XI of the ASME Code. Classify flaw as either:

Inside Surface Outside Surface Subsurface X

4) Determine Vessel Wall Geometiv. If the flaw is classified as subsurface or outside surface, input 0 for clad thickness, else enter the analysis value for clad thickness as listed in Table A-1 of Appendix A for the specified weld region.

Cladding Thickness, telad = 0 (in)

Low Alloy Steel Thickness, tLAS = __5.875 (in)

Total thickness, t=telad + tLAS =

__5.875 _in)

(

l 8 Reference wcld no. V3b ,1-234B/ Figure 1 2 " Rollout drawing", page 10 of the Handbook (Attachment 3.1) .

Reference Dwg. no. CPR 10ll and exam data sheet 10-02 of attached weld examination sheets.

66 L

t

)

GENuclear Energy GENE-523-008-0194 Rev.1 DRF # 137-0010-7 (Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID: VLC-BB-2.

Indications 10-002/10-003/10-004 ---

Doc , NEDC 48-449 AT TACHMEtii - 3*7 PAGE 0 __0F - M .

1

5) Eize Flaw. Calculate flaw depth, including any portion of the flaw extending into l

the cladding.

Surface Flaws: Subsurface Flaws:

Flaw Depth, a = N/A (in) Flaw Depth,2a = 0.356_ (in)

Flaw Length, L = N/A (in) Half Depth, a = _ _0.178__ (in)

Flaw Length, L = _3.25_ (in) 1 l

I

6) Calculate Aspect Ratio of Flaw.

Flaw Aspect Ratio, a/L = 0.055 l

7) IWB-3500 Flaw Evaluation. For the given a/L aspect ratio, determine the allowable flaw depth, a (surface) and 2a (subsurface), in accordance with IWB-3510 of the Code and record the value below. If the flaw depth recorded in I step 5 is below the allowable value, check the box " Acceptable per IWB-3500" below. Otherwise, check the box " Unacceptable per IWB-3500" and continue to !

step 8.

Inside Surface Flaw:

IWB-3500 Allowable Depth = a = __N/A (in)

Outside Surface Flaw (vessel flange regions only):

IWB-3500 Allowable Depth = a = __ N/A _____ (in)

Subsurface Flaw:

IWB-3500 Allowable Depth = 2a = _0.262 (in) < 0.356 (in)

, 67

F 1

GENuclear Energy GENE-523-008-0194 Rev.i DRF S 137-0010-7 (Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID: VLC-BB-2.

Indications 10-002/10-003/10-004 -

DOC

N D i8-O YO ATTACHMENT - .T .7 _l PAGE- 7 0F -- ZC ACCEPTABILITY:

Acceptable per IWB-3500.

X Unacceptable per IWB-3500.

8) IWB-3600 Flaw Evaluation. Record the appropriate flaw acceptance diagram l Figure # from Section 3.0. Record the allowable flaw depth, a or 2a, from the appropriate curve for the specified oriertation. If the flaw depth recorded in step 5 is below the allowable value, check the box " Acceptable per IWB-3600" ,

below. Otherwise, check the box " Unacceptable per IWB-3600", and proceed to ) l step 9.

NOTE: Flaw specific analysis would be required if outside surface flaws were found in any region below the vessel flange.

Figure # 3-8 (IWIl-3600 Axial Flaw Eval. Curve)

Inside Surface Flaw:

IWB-3600 Allowable Depth = a = N/A (in)

Outside Surface Flaw (vessel flange regions only):

IWB-3600 Allowable Depth = a = N/A (in)

Subsurface Flaw:

IWB-3600 Allowable Depth = 2a = 1.75 (in) > 0.356 (in)

ACCEPTABILITY:

X Acceptable per IWB-3600. (for 16 EFPY) l Unacceptable per IWB-3600.

68

f-r-

GE Nuclear Energy GENE-523 008-0194 Rev.1 DRF M 137-0010-7 (Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID: VLC-BB-2.

Indications 10-002/10-003/10-004 ooc , NEpc 96 ok ATTACHMENT - Je7 PAGE A 0F N

9) From figure identified above, record the 1/3 wall thickness limit below. If flaw depth is below 1/3 limit, flaw removal is acceptable. Otherwise, weld repair is necessary.

1/3 Limit = ___N/A (in)

From step 5 above:

Flaw depth = a = N/A __ (surface \

2a = ____ N/A __ (subsunuce)

Flaw excavation depth < 1/3 Limit: Flaw removal acceptable (No weld repair)

Flaw excavation depth > 1/3 Limit: Weld repair required l

l j' 69

p GENuclear Energ GENE-523-008-0194 Rev.)

DRF # 137-0010-7 000 a N E M M - 0 4 8 ATTACHMENT - 37 PAGE 9 0F- M COOPER NUCLEAR STATION FLAW EVALUATION WORKSHEET Flaw ID: VLC-BB-2 e . Indications 10-010/10-011 *

1) Determine Region and Orientation of Flaw The weld region should be identified by the nearest weld. The orientation is either [A]xial or [C]ircumferential. If the ;

flaw is at ajunction between two welds, the region with the more limiting acceptance criteria should be conservatively used.

l Region: V3 Orientation: Axial l i

2) Sketch Flaw Geometry. Use the attached flaw sketch to draw the flaw.

3) Classify Flaw. Combine flaws in close proximity to other flaws and to the surface per the proximity rule ofIWA-3300,Section XI of the ASME Code. Classify )

1 flaw as either:

l Inside Surface Outside Surface Subsurface X !

I l

Determine Vessel Wall Geometry. If the flaw is classified as subsurface or 4) outside surface, input 0 for clad thickness, else enter the analysis value for clad thickness as listed in Table A-1 of Appendix A for the specified weld region.

Cladding Thickness, tclad = 0 (in) l Low Alloy Steel Thickness, tLAS = _5.875 (in) l Total thickness, t=tc lad + tLAS = _ _ _5.875_ (in) l e Reference weld no. V3b ,1 234B/ Figure 1 2 " Rollout drawing", page 10 of the Handbook (Attachment 3.1) .

Reference Dwg. no. CPR 101I and exam data sheet 1042 of attached weld examination sheets.

f

. 66

I l l

GEhuclear Energy GENE-523-008-Ol94 Rev.I DRF H 137-0010-7 l t

(Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID: VLC-BB-2.

Indications 10-010/10-011 ooce llE N 98 o VB AMACHMENT _ _T. / l PAGE //") _ of _ 'l d

5) Size Flaw. Calculate flaw depth, including any portion of the flaw extending into I

the cladding.

l Surface Flaws: Subsurface Flaws:

Flaw Depth, a = N/A (in) Flaw Depth,2a = _ _0.39 __ (in)

Flaw Length, L = N/A (in) Half Depth, a = __0.195 _ (in)

Flaw Length, L = 3.25 _ (in)

6) Calculate Aspect Ratio of Flaw.

Flaw Aspect Ratio,'a/L = _0.060

7) IWB-3500 Flaw Evaluation, For the given a/L aspect ratio, determine the allowable flaw depth, a (surface) and 2a (subsurface), in accordance with IWB-3510 of the Code and record the value below. If the flaw depth recorded in step 5 is below the allowable value, check the box " Acceptable per IWB-3500" below. Otherwise, check the box " Unacceptable per IWB-3500" and continue to step 8.

Inside Surface Flaw:

IWB-3500 Allowable Depth = a = N/A (in)

Outside Surface Flaw (vessel flange regions only):

IWB-3500 Allowable Depth = a = N/A _ (in) i Subsurface Flaw:

IWB-3500 Allowable Depth = 2a = 0.306_ (in) < 0.39 (in) l

! 67

GE Nuclear Energpr OSyg.333.gog.oj94 p,y, y DRF # 137-0010-7 l

(Cooper Nuclear Station Fla,v Evaluation Worksheet cont'd) Flaw ID:_ VLC-BB-2.

Indications 10-010/10-011 00c o NGWO-OA0 ATTACHMENT 3-7 PAGE // OF M l ACCEPTABILITY:

i Acceptable per IWB-3500.

l ._X_ Unacceptable per IWB-3500.

8) IWB-3600 Flaw Evaluation. Record the appropriate flaw acceptance diagram l

Figure # from Section 3.0. Record the allowable flaw depth, a or 2a, from the appropriate curve for the specified orientation. If the flaw depth recorded in step 5 is below the allowable value, check the box " Acceptable per IWB-3600" l below. Otherwise, check the box " Unacceptable per IWB-3600", and proceed to step 9.

NOTE: Flaw specific analysis would be required if outside surface flaws were fotmd in any region below the vessel flange.

Figure # 3-8 (IWB-3600 Axial Flaw Eval. Curve)

I Inside Surface Flaw:

IWB-3600 Allowable Depth = a = ___ N/A (in)

Outside Surface Flaw (vessel flange regions only):

IWB 3600 Allowable Depth = a = N/A (in)

Subsurface Flaw:

1 IWB-3600 Allowable Depth = 2a = 1.75 (in) > 0.39 (in)

ACCEPTABILITY:

X Acceptable per IWB-3600. (for 16 EFPY)

Unacceptable per IWB-3600.

i 68

l GE Nuclear Energy GENE-523-008-0194 Rev.1 DRF H 137-0010-7 (Cooper Nuclear Station Flaw Evaluation Worksheet cont'd) Flaw ID: VLC-BB-2.

Indications 10-010/10-011 occnN W 98-dE0 -

ATTACHMENT 77 PAGE /1 0F M

9) From figure identified above, record the 1/3 wall thickness limit below. If flaw '

depth is below 1/3 limit, flaw removal is acceptable. Otherwise, weld repair is ;

necessary. I 1/3 Limit = N/A (in) l From step 5 above:

Flaw depth = a = N/A _ (surface) 2a = N/A (subsurface)

Flaw excavation depth < 1/3 Limit: Flaw removal acceptable (No weld repair)

Flaw excavation depth > 1/3 Limit: Weld repair required I i

l 1

I I

69 j

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DOC,MEM M-dd l AT TACHMENT 37 FAGE /M OF M AzimuthalLocation 0

30 60 90 120 150 180 210 240 270 300 330 I

. . . I i . . . . . . . .

VFW Weld V4a Weld V4b Weld V4e Weld

[2 234A [2-234B [2-234C

}{34 Weld l

[V3a 1-234A Weld [V3b 1-234B Weld [V3c 1-234C Weld 1I23 Weld V2a Weld V2c Weld i g'

[1-233A Beltline Region

[V2b 12338 Weld [1-233C

...... ........................[......................... . . .. . . .............H.

1. 2..W...e.l.d...... ..... .. . . .. .

....................................... I l

Vla Weld Vib Weld Vic Weld 1

[2-233A [2-233b [2-233C 1 l

FIGURE l-2 Cooper Nuclear Station Partial RPV" Rollout" Drawing i

. 10

00c o NGDt.-je ot/f '

Al l Ai,HMtNi J + /

" GERIS 2000 INVESSEL REeORT NO.:

GE Nuclear Energne " ' * -

SUMMARY

SHEET PROJECT: __CODEERIJNIT 1.RE18 PROCEDURE: __GE UIJD0V2 REV._N/A. FRR: _1BBE3-001_

_N/A RA SYSTEM: _REACIORPRESSURE YESSEL PROCEDURE: _GE-UIJ01V2 REV N/A FRR: __N/A

_ N/A WELD NO.: _VLC-BB-2 N/A PROCEDURE: 11I-CNS-300V3 REV._Q_ FRR: 1G00@1.

CONFIGURATION: _1ONGIIUDINAL _ N/A __

_EA DATA SHEET NO.(S): Examination Data Sheets 1001 thru 10-04, Indication Data Sheets 10401 thru 10-059, Indication Screen Prints10-001 thru 10459, Indication Plots, Indication Evaluation Sheets, Exam Patch Location Map with Coverage Data and GERIS 2000 invessel Setup Records.

Manual Calibration Sheets RC-022 , RC-023, RC-024 , Manual Data Sheet RD 022 Weld VLC-Be-2 was examined utilizing 0* iongitudinal wave,45* shear wave,60* shear wave and 70* refracted longitudinal (RL) wave techniques.

The ASME Section XI required examincon volume was examined with the GERIS 2000 invessel System from the RPV inside surface.

T' e automated examination coverage was calculated to be 73.8%.

Areas limited to the GERIS 2000 were examined manually utilizing O' longitudinal wave,45' shear wave, and 60* shear wave techniques from the RPV outside surface. Access to the OD surface was drom the N4C feedwater nozzle bloshield window.

The manual examination was not limited.

The manual examination coverage was calculated to be 36.3% (25% applied).

The total examination coverage was calculated to be 98.8%.

Forty nino (49) flaw indications were recorded with the GERIS 2000 and evaluated as be'ng acceptable to the requirements of IWP 3500-1.

Indic' i'ons 10402,10-003 and 10404 were combined in accordance with IWA-3330 and evaluated as being unacceptable. g indications10-010 and 10-011 were combinod in accordance with IWA-3330 and evaluated as being unacceptable. /(f g.o er2.

Indications10-048 and 10-049 were combined in accordance with lWA-3330 and evaluat3d as being unacceptable.

The unacceptable flaw indications were plotted to the region of the weld root and are attributed to slag remaining from insufficient back gouging during fabrication.

Three (3) geometric indications from the outside surface were recorded.

Seven (7) geometric indications from an outside surface attachment (insulation support bracket @120* azimuth) were recorded.

)

Previous data was reviewed prior to the completion of this summary. The areas contelning the unacceptable indications were not accessible during the previous examinations.

Examination surface painted Reference Letter to File " Ultrasonic inspection Through Painted Surfaces".

The examination results are ut, acceptable per the requirements of ASME Section XI,1989 Edition No Addenda.

W:fI* ?' _

f ?_ _!f S h._ - -

SUMMARY

Y; LE' VEL DATE NPPD NDE REVIEW eY: DATE

11. ] $ - . . - - -

pact 3 or

(. ) ..naw e,; e o- s en ea .-w u,; o- -

r DOCO Y W $b"OY0 AT TAChMENT .L7 PAGE -- /b 0F M GERIS 2000 Indication G, , GENuclearEnergy Evaluation Data Sheet Project: Cooper Nuclear Power Station Exam Data Sheet: 10-02 WeldlD: VLC-BB-2 Ind. Data Sheet: 10-002/10-003/10 004 Patch 10: V3B-01 Indication : 2->3->4 Flaw ThruwallDimension = 0.356 *T* nominal = 5.88 Flaw Length *l* = 3.25 *T* measured a N/A Surface Separation *S* = 1.40 Clad "T" nominal = 0.31 i a l ASME Section XI,1989 Edition, No Addenda l TABLE IWB-3510-1 for 4" and Greater all Surface % Subsurface % Surface % Subsurface %

0.00 1.9 2.0 - -

0.05 2.0 2.2 2.02 2.23 Y 0.10 2.2 2.5 - -

0.15 2.5 2.9 - -

0.20 2.8 3.3 - -

0.25 3.3 3.8 - -

0.30 3.8 4.4 - -

0.35 4.4 5.1 - -

0.40 5.0 5.8 - -

0.45 5.1 6.7 - -

0.50 5.2 7.6 - -

Allowed Allowed 2.02 2.23 l

a= 0.178 all value = 0.055 Y= 1.000 Flaw is Subsurface Allowed alt = 2.23 %

a/t = 3.03 %

Flaw is unacceptable by Table IWB-3510-1.

Comments : Axial.

Combined flaws10-002,10-003 and 10-034 per IWA-3330.

Analyst: Reviewed By:

Level: Date: Level: Date:

r 1 l

DOCeNEO6Y8-640 l AT TACWE NT - 3 . ~7 PAGE _ / 7 Of - 93 GERIS 2000 Indication GE Nuclear Energy Evaluation Data Sheet l k l

i Profect: Cooper Nuclear Power Station Exam Data Sheet: 10-02 WeldlD: VLC-BB-2 Ind. Data Sheet: 10-010/10-011 Patch ID : V38-01 Indication : 10->11 Flaw ThruwallDimension = 0.39 *T* nominal = 5.88 Flaw Length *l* = 3.25 "T* measured = 5.90 Surface Separation *S* = 2.10 Clad *T* nominal = 0.31 ASME Section XI,1989 Edition, No Addenda TABLE IWB-3510-1 for 4" and Greater all Surface % Subsurface % Surface % Subsurface %

0.00 1.9 2.0 - -

0.05 2.0 2.2 2.04 2.26 Y 0.10 2.2 2.5 - -

0.15 2.5 2.9 - -

0.20 2.8 3.3 - -

0.25 3.3 3.8 - -

0.30 3.8 4.4 - -

0.35 4.4 5.1 - -

0.40 5.0 5.8 - -

0.45 5.1 6.7 - -

0.50 5.2 7.6 - -

Allowed Allowed 2.04 2.26 a= 0.195 all value = 0.060 Y= 1.000 Flaw is Subsurface Allowed alt = 2.26 %

alt = 3.31 %

Flaw is unacceptable by Table IWB-3510-1.

Comments : Axial.

Combined flaws10-010 and 10-011 per IWA-3330.

Analyst: Reviewed By:

Leve!: Date: Leve:. Date:

DOC # W 98 d f f 1 ATTACHMENT __ .L 7 PAGE- /A _ Or _ M

,i GERIS 2000 Indication GE Nuclear Energy Evaluation Data Sheet Project : Cooper Nuclear Power Station Exam Data Sheet: 10-04 WeldID : VLC-BB-2 Ind. Data Sheet: 10-048/10-049 PatchID: V38-03 Indication: 48->49 Flaw ThruwallDimension = 0.40 *T* nominal = 5.88 Flaw Length *I" = 1.25 "T' measured = N/A Surface Separation 'S" = 2.70 Clad *T* nominal = 0.31 ASME Section XI,1989 Edit lon, No Addenda TABLE IWB-35101 for 4" and Greater !

all Surface % Subsurface % Surface % Subsurface %

0.00 1.9 2.0 - -

0.05 2.0 2.2 - -

0.10 2.2 2.5 - -

0.15 2.5 2.9 2.56 2.98 Y 0.20 2.8 3.3 - -

0.25 3.3 3.8 - -

0.30 3.8 4.4 - -

0.35 4.4 5.1 - -

0.40 5.0 5.8 - -

0.45 5.1 6.7 - -

0.50 5.2 7.6 - -

Allowed Allowed 2.56 2.98 l

a= 0.200 all value = 0.160 Y= 1.000 l 1

l Flaw is Subsurface Allowed a/t = 2.98 %

alt = 3.40%

)

Flaw is unacceptable by Table IWB-3510-1.

j Comments : Axial. !

Combined flaws10-048 and 10-049 per IWA-3330.

Analyst: Reviewed By:

Level: Date. Level: Date:

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Text

1.0 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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Description:

8.0 CONCLUSION

SUMMARY

SUMMARY

1.0 INTRODUCTION

SUMMARY

5.0 REFERENCES

1.0 INTRODUCTION

: :

SUMMARY

5.0 REFERENCES

27,600 ksi modulus at 195 F AT

Reference:

Dear Mike:

Subject:

Dear Mr. Bacha:

Subject:

Dear Mike:

Subject:

DearMike:

Subject:

Dear Ali:

Subject:

References:

Dear Ali:

Subject:

References:

Dear Ali:

Subject:

References:

Dear Ali:

Reference:

Dear Mike:

Subject:

Dear Mr. Bacha:

Subject:

Dear Mr. Bacha:

Subject:

Subject:

Subject:

Subject:

Subject:

Purpose:

References:

Conclusion:

Conclusion:

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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