NL-08-053, 10 CFR 50.55a Request RR-CRV-75 - Relief from Examinations of Component Welds with Less than Essentially 100% Examination Coverage for Third-Ten Year Inservice Inspection Interval Closeout

10 CFR 50.55a Request RR-CRV-75 - Relief from Examinations of Component Welds with Less than Essentially 100% Examination Coverage for Third-Ten Year Inservice Inspection Interval Closeout

ML080920717
Person / Time
Site:	Indian Point
Issue date:	03/26/2008
From:	Robert Walpole Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-08-053, RR-CRV-75
Download: ML080920717 (74)

Text

Indian Point Energy Center 450 Broadway, GSB En TO?' P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 734-6700 R. Walpole (914) 734-6710 March 26, 2008 Re: Indian Point Unit 2 Docket No. 50-247 NL-08-053 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

10 CFR 50.55a Request RR-CRV Relief from Examinations of Component Welds with Less Than Essentially 100% Examination Coverage for Third-Ten Year Inservice Inspection Interval Closeout

Reference:

1. Con Edison Letter NL-94-006, "Third Ten-Year Interval Inservice Inspection Program," dated January 24, 1994.

2. ASME Section Xl, Code Case N-460, "Alternative Examination Coverage for Class 1 and Class 2 Welds, Section Xl, Division 1"

Dear Sir or Madam:

Indian Point Unit 2 (IP2) Third ten-year inservice inspection interval ended on March 1, 2007. During the Third interval, IP2 completed the required in-service examinations in accordance with the program plan (Reference 1); except, certain components could not fully meet the volumetric examination requirements stipulated in the ASME Section XI Code, 1989 Code Edition, with No Addenda, including the clarifications provided in the ASME Code Case N-460 (Reference 2). Entergy has determined that conformance with the code requirement of essentially 100% coverage of weld volume or area examined was impractical due to various constraints and limitations. Accordingly, pursuant to 10 CFR 50.55a (g)(5)(iii), Entergy submits the attached IP2 Relief Request, RR-CRV-75, for NRC review and approval. Relief Request, RR-CRV-75, proposes alternatives where the requirement of "essentially 100%" volumetric examination was not feasible due to construction limitations, obstructions, accessibility, and examination techniques.

The alternatives and justifications are explained in the attached relief request providing a list of components which requires relief pursuant to 10 CFR 50.55a. The alternatives and justifications provide acceptable level of quality and safety and will not adversely impact the health and safety of the public.

ALI 7 AJ~

Indian Point Unit2 Docket No. 50-247 NL-08-053 Page 2 of 2 There are no new commitments identified in this submittal. If you have any questions or require additional information, please contact Mr. R Walpole at (914) 734-6710.

Sincerely, R Walpole Licensing Manager Indian Point Energy Center

Attachment:

A. 10 CFR 50.55a Relief Request RR-CRV-75 cc: Mr. John P. Boska, Senior Project Manager, NRC NRR DORL Mr. Samuel J. Collins, Regional Administrator, NRC Region 1 NRC Resident Inspector, IP2 Mr. Paul D. Tonko, President NYSERDA Mr. Paul Eddy, New York State Dept. of Public Service

Attachment A to NL-08-053 10 CFR 50.55A RELIEF REQUEST RR-CRV-75 RELIEF FROM EXAMINATIONS OF COMPONENT WELDS WITH LESS THAN ESSENTIALLY 100% EXAMINATION COVERAGE FOR THIRD-TEN YEAR INSERVICE INSPECTION INTERVAL CLOSEOUT ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 DOCKET NO 50-247

Attachment A Docket No. 50-247 NL-08-053 Page 1 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Proposed Alternative In Accordance with 10 CFR 50.55a(g)(5)(iii)

Inservice Inspection Impracticability

1. ASME Code Comoonent(s) Affected Code Class: Class 1 (Quality Group A)

Examination Categories: B-A, B-D, and R-A Item Numbers: B1.22, B1.40, B3.120, and R1.20

2. Applicable Code Edition and Addenda

The Code of Record for Indian Point Unit 2 Inservice Inspection Third Ten-Year Interval is the ASME Section XI Code, 1989 Edition, No Addenda.

3. Anolicable Code Reauirements IWB-2500, states in part, "Components shall be examined and tested as specified in Table IWB-2500-1. The method of examination for the components and parts of the pressure retaining boundaries shall comply with those tabulated in Table IWB-2500-1, except where alternate examination methods are used that meet the requirements of IWA-2240". Table IWB-2500-1 requires that a volumetric and/or surface examination be performed on specified components based on Code Category and Item Number. The applicable examination area or volume and method required are shown below from Table IWB-2500-1:

<ASME~ [ ASSME Section XlExamn Examination ~7~ASME Item~ Requirements/ Figure NDE Examination

-Category . Number Number Method B-A B 1.22 IWB-2500-3 Volumetric B-A B1.40 IWB-2500-5 Volumetric & Surface B-D B3.120 IWB-2500-7 Volumetric R-A R1.20 IWB-2500-8 Volumetric

4. Impracticality of Compliance Relief is requested from performing a complete coverage examination of the entire volume of area required. Entire volume or area required is defined by ASME Section Xl Code Case N-460 titled "Alternative Examination Coverage for Class and Class Welds,Section XI, Division 1 ."

Code Case N-460 states in part, "...when the entire examination volume or area cannot be examined ... a reduction in examination coverage.. .may be accepted provided the reduction in coverage for that weld is less than 10%."

The NRC through, Information Notice 98-42 titled "Implementation of 10CFR50.55a(g) Inservice Inspection Requirements," termed the reduction in

Attachment A Docket No. 50-247 NL-08-053 Page 2 of 71 10 CFR 50.55a Relief Request RR-CRV-75 coverage of less than 10% to be essentially 100 percent." Information Notice 98-42 states, in part, "The NRC has adopted and has further refined the definition of

'essentially 100% percent' to mean 'greater than 90 percent'...has been applied to all examinations of welds or other areas required by ASME Section Xl."

The construction permit for Indian Point Unit 2 was issued on October 14, 1966. At that time, the ASME Boiler and Pressure Vessel Code covered fabrication of only nuclear vessels. Piping, pumps, and valves were built primarily to the rules of USAS B31.1.0-1955, Power Piping. The IP2 systems and components were designed and fabricated before the examination requirements of ASME Section XI were formalized and published. Therefore, IP2 was not specifically designed to meet the requirements of ASME Section Xl and full compliance is not feasible or practical within the limits of the current plant design.

10 CFR 50.55a recognizes the limitations to inservice inspection of components in accordance with Section XI of the ASME Code that are imposed due to early plants' design and construction, as follows:

10 CFR 50.55a(g)(5)(iii): If the licensee has determined that conformance with certain code requirements is impracticalfor its facility, the licensee shall notify the Commission and submit, as specified in § 50.4, information to support the determinations.

Accordingly, pursuant to 10 CFR 50.55a(g)(5)(iii), Entergy has determined that conformance with the code requirement of essentially 100% coverage of weld volume or area examined was impractical due to various constraints and limitations as stated above. Entergy requests NRC approval of the proposed alternative as stated below.

Relief is requested from performing an examination of "essentially 100% of the required volume or area as applicable for the identified components in Table 1 below.

ASME Code Category and Item Numbers where previous relief requests were granted for coverage limitations not identified in Table 1 are as follows:

Class 1 (Quality Group A)

ASME Code Category B-A, "Pressure Retaining Welds in Reactor Vessel Head Welds," (Item Nos. B1.21 and B1.22), are not listed in Table 1. Relief was submitted and granted for coverage limitations for meridional welds RVHM-1, RVHM-3, &

RVHM-5 in the following approved relief request:

• Relief Request RR-06, Rev. 1, Reactor closure head circumferential welds. NRC Relief Granted, Date 6-3-97 3rd Int. TAC No. M88559.

ASME Code Category B-D and B-J, "Surface examination of Reactor Vessel Nozzle to Safe End Welds" (Item Nos. B5.10 and B9.11), are not listed in Table 1. Relief was submitted and granted for impracticality of examination in the following approved relief request:

Attachment A Docket No. 50-247 NL-08-053 Page 3 of 71 10 CFR 50.55a Relief Request RR-CRV-75

• Relief Request RR-1 1, Rev. 1, Surface examination of Reactor Vessel Nozzle to Safe End Welds. NRC Relief Granted, Date 6-3-97 3rd Int. TAC No. M88559.

ASME Code Category B-B, ."Pressure Retaining Welds in Vessels Other than Reactor Vessels," (Item Nos. B2.11 and B2.12), Pressurizer Shell-to-Head welds, not listed in Table 1. Relief was submitted and granted for coverage in the following approved relief request:

• Relief Request RR-07, Pressurizer shell to head, circumferential and longitudinal welds. NRC Relief Granted, Date 6-3-97, TAC No. M88559.

ASME Code Category B-B, "Pressure Retaining Welds in Vessels Other than Reactor Vessels," and Category B-D, "Full Penetration Welds of Nozzles in Vessels," (Item Nos. B2.51, B2.80, B3.150, and B3.160), Regenerative Heat Exchanger vessel welds, not listed in Table 1. Relief was submitted and granted for coverage in the following approved relief requests:

" Relief Request RR-08, Rev. 2, Regenerative Heat Exchanger vessel welds and inner radius sections. ASME Cat: B-B, B-D. NRC Relief Granted, Date 6-3-97, TAC No. M88559.

• Relief Request RR-60, Alternative to eliminate ASME Cat. B-B and B-D weld examinations on regenerative heat exchanger. NRC Relief Granted, Date 1/17/2003, TAC No. MB5834.

ASME Code Category B-D "Pressure Retaining Welds in Vessels Other than Reactor Vessels," (Item No. B3.120), Pressurizer Nozzle Inside Radius, are not listed in Table 1. Relief was submitted and granted for coverage in the following approved relief request:

• Relief Request RR-09, Rev. 1, Pressurizer Inside Radius sections.

ASME Cat B-D. NRC Relief Granted, Date 6-3-97, TAC No. M88559.

ASME Code Category B-J "Pressure Retaining Welds in Piping" (Item No. B9.11 and B9.12), are not listed in Table 1. Relief was submitted and granted for coverage in the following approved relief requests:

" Relief Request RR-27, Reactor coolant piping, longitudinal welds. ASME Cat: B-J. NRC Relief Granted, Date 6-3-97, TAC No. M88559.

• Relief Request RR-51, Line 351, Weld 351-2, >90% examination coverage. NRC Relief Granted, Date 11/7/2000, TAC No. MA6909.

ASME Code Category B-K, "Integral Attachments for Pumps," (Item No. B10.10),

are not listed in Table 1. Relief was submitted and granted for coverage in the following approved relief request:

• Relief Request RR-38, Documentation of less than 90% coverage, RCP integral welded attachments. NRC Relief Granted, Date 3-21-00, TAC No. MA5918.

Attachment A Docket No. 50-247 NL-08-053 Page 4 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Class 2 (Quality Group B)

ASME Code Categories C-A, C-B, C-C, (Item Nos. C1.10, C1.20, C2.31, C3.10), are not listed in Table 1. Relief was submitted and granted for coverage in the following approved relief requests:

• Relief Request RR-16, Residual Heat Exchange nozzle to vessel and integral attachment welds. NRC Relief Granted, Date 6-3-97, TAC No. M88559.

• Relief Request RR-52, RHRHX, Welds RHX C22-1 & C22-2. NRC Relief Granted, Date 11/7/2000, TAC No. MA6909.

5. Burden Caused By Compliance Compliance with the examination coverage requirements of ASME Section Xl would require extensive engineering, modification, redesign, or replacement of components where geometry is inherent to the component design.

6. Proposed Alternative and Basis for Use Proposed Alternative A) The components listed in Table 1 have already been examined by the available methods to the maximum extent practical. No additional volumetric or surface examinations will be performed on the components for the 3rd Inservice Inspection Interval.

B) A visual inspection (VT-2) was performed by VT-2 qualified operators on the subject components during the system pressure tests (with no leakage detected) as required by code category B-P (each refueling outage).

Basis for Use The Reactor Pressure Vessel and Pressurizer, and class 1 piping were designed and fabricated to Codes in effect during the late 1960's. These Codes did not require that there be full access for Inservice inspection, as was required by later Codes.

ASME Code Category B-A, Item No. B13.22, Meridional Welds-RVHM-2, RVHM-4, &

RVHM-6: The RV closure head peel segment to disc circumferential weld and portions of the intersecting meridional welds are completely enclosed within the pattern of CRDM penetrations inside the shroud and, as such, are completely Inaccessible for volumetric examination as would be required by IWB-2500. (See Attachment 1)

ASME Code Category B-A, Item No. B1.40, RV Head Flange Circ Weld RVHC-2:

The Reactor Vessel Flang-to-Head circ weld has limited accessibility due to the contour of the flange head and the interference of the lifting lug. (See Attachment 2)

ASME Code Category B-D, Item No. B3.120, Pressurizer Spray & Surge Nozzle Inside Radius Welds PZRN-1 and PZRN-6: The nozzles of the pressurizer are cast with the vessel heads. The as-cast surface of the heads combined with the geometry of this area makes ultrasonic examination of the inner radii impractical.

Attachment A Docket No. 50-247 NL-08-053 Page 5 of 71 10 CFR 50.55a Relief Request RR-CRV-75 The spray nozzle inner radius (PZRN-1) is also covered by an array of nozzle heads.

The surge nozzle inner radius (PZRN-6) is also inaccessible due to the area being covered by a retaining basket. (See Attachment 3)

ASME Code Category R-A, Item No. R1.20, RCS Pipe to Safe-End Circ Welds RCC-21-1, RCC-22-1, RCC-23-1, and RCC-24-1: These reactor vessel pipe to safe end had limitations due to the tapered area of the weld overlay on the ID of the pipe.

(See Attachment 4)

Physical obstructions imposed by design, geometry and materials of construction are typical of vessel appurtenances, structural and component support members, adjacent component weldments in close proximity, and unique component configurations.

As a minimum, all components received the required examination(s) to the extent practical with regard to the limited or lack of access available. The examinations conducted confirmed satisfactory results evidencing no unacceptable flaws present, even though "essentially 100%" coverage was not attained. IP2 has concluded that if any active degradation mechanisms were to exist in the subject welds, those degradations would have been identified in the examinations performed.

For volumetric (ultrasonic) examinations, the transition to PDI examinations varied the means of coverage determination based on code requirements. Earlier Code examination coverage was determined by a simple average of the examination scans performed. Later examinations performed per PDI requirements identify coverage based on a simple average of the scans per the requirement of PDI and CFR. The attachments within identify the means of determining examination coverage and are based on the requirements at the time the examinations were performed.

To summarize, IP2 has examined all components in the 3rd 10-Yr Interval ISI Program and associated augmented programs to the maximum extent possible given the inspection limitations discussed above.

When the IP2 ISI Program is viewed in total, the overall degree of coverage obtained is still greater than 90%, i.e. essentially 100%. For this and the other reasons detailed in this request, Entergy believes that the limited coverage obtained on the components listed in Table 1 is not significant and will provide an adequate level of quality and safety for examination of the affected welds, and will not adversely impact the health and safety of the public.

.7. Duration of Proposed Alternative Relief is requested for the third ten-year interval of the Inservice Inspection Program for Indian Point Unit 2, which began July 1, 1995 and concluded March, 2007.

Attachment A Docket No. 50-247 NL-08-053 Page 6 of 71 10 CFR 50.55a Relief Request RR-CRV-75

8. Attachment Indian Point Unit 2 Third Interval ISI program datasheets for examinations with less than "Essentially 100%" coverage are attached (Attachments 1 - 4)

9. References I. NRC Information Notice 98-42, "Implementation of 10CFR50.55a(g) Inservice Inspection Requirements."

2. ASME Section XI Code Case N-460, "Alternative Examination Coverage for Class 1 and Class 2 Welds."

3. ASME Section XI Code Case N-613-1, "Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-D, Item No's. B3.10 and B3.90, Reactor Nozzle-To-Vessel Welds."

4. Relief Request RR-63, Risk-Informed In-service Inspection Program. ASME Cat B-F, B-J. Ref: EPRI TR-112657. NRC Relief Granted, Date 03/19/2004, TAC No. MC0624.

5. Relief Request RR-67, Code Case N-613-1: Alternative for reduced (UT) exam for Full Penetration Nozzles in Vessels, ASME Category B-D, item No's. B3.10 and B3.90. NRC Relief Granted, Date 7/7/2004, TAC No. MC1698.

6. Relief Request RR-73, Extend the Third 10-Yr Reactor Vessel Weld Examination Inservice Inspection one refueling cycle; including ASME Code Categories B-A Item B1.12, Item B1.21, Item B1.22, Item B1.30, ASME Code Category B-D, Item B3.90, & Item B3.100. NRC granted via SER, TAC No. MC7306, dated February 22, 2006.

Attachment A Docket No. 50-247 NL-08-053 Page 7 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Table 1 Summary of Components with Less than "Essentially 100% Code Required Coverage"

~ASME' I to,,Pretg

. *C.... ASME , ,Component ... Component -i.. NDE Condition Limeiting

]e N RCS ID[ 1 MrV TMethod CodI ergnaccesibedut e o tecipii y 'Achievedl UppRCSern elds (UoMT) CRDM Penetration 85% Attachment1 B-A BRVHM-2 U and Shroud area.

B Inaccessible dueato RV Upper Head Volumetric B-A B1.22 RCS RVHM-4 d CRDMn Penetration 85% Attachment 1 I*Merio

______ jInaccessible onal vvelds

_____________ (UT) and Shroud area.

due to B-ARCSB i .40RVHC-2 RV Uppe Upper Head Hea Volumetric j1One due tosided Flange exam and 50% Atcmn2 B-AB1.0CS VH-2 lane eld */(UT/MT) *: Head shroud i* *_ interference B-D B3.120 2 RCS PZRN-1 Ienner (t f nozzerhed ad baAttachment 0% 3 J_ __ __ 1 Radius Weld (UT)

RI Ups er Surg 1 Volumetric Covered by a I B-D B3.120 RCS U PZRN-6 Noz. Inner Radius I retaining basket 0% Attachment 3

__g Weld e (UT'...')

RCS Pipe to Safe-, . Limitation dueto R-DA B3Ri

.201 RCS RCC 21-1 End Circ Weld J R@20 0 VolumTric (UT) orerainong of the 88% Attachment 4 CotaeredareathekeD 1

I metric

Attachment A Docket No. 50-247 NL-08-053 Page 8 of 71 10 CFR 50.55a Relief Request RR-CRV-75 RRSMEE tapero d of t Co ~ ASMVE Component ~Component~ NDE Cndition Limiting- erea o Catecioryl Ie N. tem ID __

Description Method ____

Coverage ID___________ ___

]RCSPipe to@Safe- 1, Limitation due to I -

R.0RC C 221 End Circ Weld Volumetrice of the 88.5% Attachment 4 i-I-R1.20 ROS RCC22-1 RO@158 0 (UT) overlay on the ID _______ __

1. n P t UtRCS Pipe to Safe- Volumpetric Lmatsionfuedt R-A Ri1.201 ii RCS RCC 23-1 End Circ Weld Tt) tapered area of the 89% Attachment 4i welds ithdtmninfe iRO@3380 verlay on the IDrcufeenialpiinwedsweeasine omleo 1RS e Circwerel c Lmitation due to R-A R1.20 RoCS RCC 24-1 Eoneusd (UT) taper area f the n 89% Attachment

____ -_____________ 0 overlay on the ID _ _____

Notes:

1. Indian Point Unit 2 received NRC approval on March 19, 2004 (TAO No. MC0624) to implement a risk-informed inspection program for Class 1 Category B-F and B-J piping welds based on the methodology detailed in EPRI Topical Report TR-1 12657, Revision B-A, and Code Case N-578 as an alternative to the requirements of the 1989 edition ASME Xl code, No addenda. The risk-informed methodology used at IP2 includes all category B-F and B-J welds in the determination of the final risk-informed inspection sample of 61 Class 1 welds. In 2006, Class 1 circumferential piping welds were assigned alternate examination category and code item numbers that were consistent with ASME Section XA Code Case N-578-1. The numbering system established in Code Case N-578-1 is similar to the one used in Code Case N-578. However, the Code Case N-578-1 numbering system is more complete and more accurately reflects the technical criteria established in EPRI Topical Report No. TR-1 12657. For these reasons, the numbering system established in Code Case N-578-1 was used instead of the one shown in Code Case N-578; and is carried forward into the fourth interval. Item number RI.20 as shown in Table 1 of RR-CRV-75 is not listed in Code Case N-578 but was included in Code Case N- 578-1 as a means to categorize elements that are not subject to a damage mechanism. Inclusion of item number R1 .20 in Table 1 has no substantive impact on relief request RR-CRV-75.

Attachment A Docket No. 50-247 NL-08-053 Page 9 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 1 Code Category B-A: Pressure Retaining Welds in Reactor Vessel Item No. B1.22 Reactor Vessel Closure & Bottom Head Welds - Meridional Components Numbers:

RVHM-2 Reactor Vessel Bottom Head Weld - Meridional RVHM-4 Reactor Vessel Bottom Head Weld - Meridional RPVM-6 Reactor Vessel Bottom Head Weld - Meridional Basis:

Meridional welds RVHM-2, RVHM-4, & RVHM-6 exam coverage are obstructed by the CRDM penetrations, RV head shroud, and flange contour. Examinations of these meridional welds were conducted to inspect as much as reasonably practical. These exams were limited by geometry or access. (See Relief Request RR-06, Rev.1, Reference 1, below)

The total meridional weld length from the RV head flange weld to the top of the head peel segment weld is -57 inches. With the head shroud cover in place, approximately 32 inches of weld length is inaccessible; therefore, only -25 inches of weld length is accessible for examination. Of this accessible weld length of -25 inches, only 85% of exam coverage was obtained, which is less then 91% coverage as stated in the NRC SER (Reference 1 below).

In summary, the reactor vessel closure head meridional welds were volumetrically examined to the extent practical. Three of the six meridional welds (RVHM-2, RVHM-4, & RVHM-6) were available for approximately 85% rather than 91% of their length for volumetric examination. Attachment 1 shows the Reactor Vessel drawing with the shroud and CRDM's covering most of the meridional welds and the References

1. Relief Request RR-06, Rev. 1, for Items B1.21 and B1.22 has been previously granted for RVHM-2, RVHM-4, RVHM-6, RVHM-1, RVHM-3, & RVHM-5 via SER, TAC No. M88559, dated June 3, 1997.

Attachment A Docket No. 50-247 NL-08-053 Page 10 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 1 RV Upper Head CRDMs/ShroudSupport (Ref. 206913)

-GQYG "W.. V4k.Atd" 0`44 FGw* A"06-REACTOR PZE55URE VE55EL NO1 A T"U C,4147ATI-W. Fall

Attachment A Docket No. 50-247 NL-08-053 Page 11 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 1 RV Upper Head Meridional Welds/CRDMs/Shroud (Ref. 206913)

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Attachment A Docket No. 50-247 NL-08-053 Page 12 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 1 RV Upper Head Meridional Weld RVHM-2 c - '.T Weld No.: RVHM2 Summary No.: 206913-RVHM2 Report No.: 04-UT097 Page 5 of f-Coverage Plot

1. The weld was scanned 100% perpendicular with a 600 RL using a Zone 1 and Zone 2 Cal.

2. < 100% coverage was achieved when scanned parallel to the Weld due to the flange at the bottom of the weld and a taper 25" up on the weld.

3. < 1% of the lower 15% of the weld was scanned in all four directions.

4. Using one sided qualified personnel, 100% of the upper 85% was covered in one direction perpendicular and one direction parallel.

5. Weld coverage of the lower 25" of the weld = 85%.

V -O

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Attachment A Docket No. 50-247 NL-08-053 Page 13 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 1 RV Upper Head Meridional Weld RVHM-4 Supplemental Report *7L 3 oW3 Report No.: 06-UT097 Page: 2 of 2 Summary No.: 206913-RVHM4 Examiner: Serth, Joseph Lvel: PDI Reviewer: Date:

Examiner: Gronewod .,PDI Site Revie. Date:

Other: NIA l: NVA ANII Review: Date:

Comments:

Coverage Plot

1. The weld was scanned 100% perpendicular with a 60" RL using a Zone 1 and Zone 2 Cal.

2. < 100% coverage was achieved when scanned parallel to the weld duo to the flange at the bottom of the weld and a taper 25" up on the weld.

3. < '% of the lower 15% of the weld was scanned In all four directions.

4. Using one sided qualified personnel. 100% of the upper 85% was covered in one direction perpendicular and one direction parallel.

5. Weld coverage of the lower 25" of the weld = 85%.

Attachment A Docket No. 50-247 NL-08-053 Page 14 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 1 RV Upper Head Meridional RVHM-6 Weld No.: RVHM6 Siumtnivr No,: 206913-RNIVHN6 Report. No.: 04-,UT084 Page (. of 6 Coverage Plot

1. The weld was scanned 1001%perpendicular with a 60' R.L using a Zone I aind Zone 2 Cal.

2. < 100"0% coverage wva achieved when an::!nred parallel Io the weld due to the iilange at the bottof) of the weld aind a taper 25" tiluon ihe weld.

3. <I (X flthe lower I5% of the weld was s anned in all four directions.

4. Using one sided qualilied pt-rsonnel,. 1(00'X of the tipper *5'ý/%was covered in one direction perpendicular aind one direction parallel.

5. Weld coverage of the lower 25" ofidie weld 85%.

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Attachment A Docket No. 50-247 NL-08-053 Page 15 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 2 RV Flange-to-Head Weld RVHC-2 Code Category B-A: Pressure Retaining Welds in Reactor Vessel Item No. B1.40 Reactor Vessel Closure Head Flange-to-Head Weld -

Circumferential Components Number:

RVHC-2 Reactor Vessel Closure Head Weld - Circumferential Basis:

Table IWB-2500-1, Examination Category B-A, Item B1.40, requires 100% volumetric and surface examination of the reactor vessel head-to-flange weld, as defined by Figure IWB-2500-5. Entergy performed the volumetric examination to the extent practical; resulting in approximately fifty percent (50%) of the weld volume was ultrasonically examined for 2/3 of the weld length in 1997. Eighty four percent (84%) coverage was obtained using updated equipment and better technique in the last 1/3 of the weld length in 2004 (See attached).

The surface examination conducted confirmed satisfactory results evidencing no unacceptable flaws present. Based on the percent of the volumetric examination completed and the Code-required surface examination, it is reasonable to conclude that degradation, if present, would have been detected.

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Attachment A Docket No. 50-247 NL-08-053 Page 17 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 2 RVHC-2 Flange to Head Cu-st-uc:ors Pag Weld

Attachment A Docket No. 50-247 NL-08-053 Page 18 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 2 RVHC-2 Flange to Head Weld Co~struc-ors Page 2 of _Z_

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)Westinghouse LIMITATION TO EXAMINATION

,LANT INDIAN POINT UNIT 2 SKETCH A206913-1

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Attachment A Docket No. 50-247 NL-08-053 Page 20 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 3 Code Category B-D: Full Penetration Welds of Nozzles in Vessels Item No. B3.120 Pressurizer Nozzle Inside Radius Section Components Number:

PZRN-1 Pressurizer Spray Nozzle Inner Radius PZRN-6 Pressurizer Surge Nozzle Inner Radius Basis During the remote visual examinations of the Pressurizer inner radius examinations conducted in May of 1997; an assessment of the ability to remote visually examine the PZRN-1 & PZRN-6 nozzles were conducted by a Level III visual examiner. The remote visual examination revealed that the thermal shield used in the manufacture of the PZRN-1 nozzle completely blocks the view of the inner radius. PZRN-1 is also covered by an array of nozzle heads. PZRN-6 is also inaccessible due to the area being covered by a retaining basket. Further, discussions that were had with the two ISI examination vendors that serviced Indian Point 2 during the Third Interval confirmed our earlier opinion that the nozzle configuration of both the PZRN-1 and PZRN-6 nozzles precluded the examination of those locations, even with the most advanced ultrasonic systems, currently in use. No coverage was obtained. (See Attachment 3).

In lieu of the Code-required volumetric examination once in ten years, the Pressurizer Spray and Surge nozzle areas, PZRN1 & PZRN-6, were visually examined (VT-2) after each refueling outage for evidence of leakage during system pressure tests performed in accordance with IWB-2500, Category B-P, and Code Case N-498. As a result no evidence of leakage were detected these areas. It is expected than any through wall defects would be detected by this examination prior to failure of the pressurizer. This is based on the expectation that the component would experience leakage before a catastrophic failure

("leak before break").

The inner radius area of pressurizer nozzles PZRN-2, PZRN-3, PZRN-4, and PZRN-5 were examined as proposed by Indian Point 2 Relief Request 9, Rev. 1, Reference 1. Nozzles PZRN-2, PZRN-4, and PZRN-5 were completed in May of 1997, and Nozzle PZRN-3 completed in April of 2006; with no recordable indications.

References:

1. NRC SER Relief Request RR-09 Rev. 1, Granted Date 6-3-97, TAC No. M88559.

2. Westinghouse Drawing No. 681J281, "General Assembly & Final Machining (1800 CU FT) Pressurizer.

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Attachment A Docket No. 50-247 NL-08-053 Page 25 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 4 Code Category R-A: Risk-Informed RV Nozzle Safe-End Weld (Formally Category B-F)

Item No. R1.20 Reactor Vessel Nozzle Safe-End Welds -

Circumferential Components Number:

RCC 21-1 RCS Pipe to Safe-End Circ Weld RO @ 2020 RCC 22-1 RCS Pipe to Safe-End Circ Weld RO @ 1580 RCC 23-1 RCS Pipe to Safe-End Circ Weld RO @ 3380 RCC 24-1 RCS Pipe to Safe-End Circ Weld RO @ 0220 Basis Examination Data Summary All automated ultrasonic examinations (implementation of Appendix VIII) were conducted in accordance with the ASME Code,Section XI, 1995 Edition with 1996 Addenda as modified by 10CFR50.55a(b)(2Xxiv, xv and xvi). Examination scan plans were prepared to the requirements of procedure PDI-ISI-254-SE Rev. 2.

Code Examinations Examinations were conducted completely by the contact technique using two WesDyne PARAGON multi-channel data acquisition systems, one interfaced to each SUPREEM robot and scanning platform. All examinations were conducted to the maximum extent practical with the access provided and within the limitation of component geometry.

Ultrasonic detection scans for the primary nozzle to safe-end and safe-end to pipe/elbow welds were examined from the ID surface using 700 L- wave transducers applied four-directionally.

Axial scans were performed at a 0.25" increment and the circumferential scans were conducted at a 0.080" incremental distance. This exam interrogated the inner 1/3 thickness volume. The safe-end to nozzle welds (RCC-21 -1, RCC-22-1, RCC-23-1, & RCC-24-1) had limitations due to the tapered area of the weld overlay on the ID (See Attachment 4). Eddy current techniques were also employed to examine the inner surfaces of the dissimilar metal and pipe/elbow welds and the adjacent examination volumes where ID geometry presents a limitation to the detection of axial flaws as defined in the PDQS for the qualified Appendix VIII technique. As a result, all areas of limitations were fully examined by the supplemental eddy current techniques with no recordable indications.

Examination Results (primary Inlet and Outlet Nozzles)

Automated ultrasonic examinations of the Indian Point Unit 2 Primary Inlet and Outlet RV Nozzle welds had no recordable indications. Supplemental eddy current examinations of the Nozzle to Safe End and Safe End to Pipe/Elbow welds from the nozzle bore yielded no recordable indications.

Attachment A Docket No. 50-247 NL-08-053 Page 26 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 4 Reactor Vessel Nozzle Safe-End Welds (A206913)

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Attachment A Docket No. 50-247 NL-08-053 Page 28 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 4 RV Nozzle Circ Weld RCC-21-1 (Datasheet)

REACTOR VESSEL WELD RESULTS

SUMMARY

PLANT NAME Indian Point Unit 2 WELD NO. RCC-21-1 COMPONENT Outlet Pipe @ 22'"

LIMITATIONS: NO j- YES KI See Coverage Breakdown Sheet RESULTS NO. OF INDICATIONS' N/A NI X STATUS N/IA RI EXAM DOCUMENTATION INDICATION DOCUMENTATION L-I PARAGON ANALYSIS LOG LI ASSESSMENT SHEET Lii PARAGON ACQUISITION LOG -,- PARAGON HARD COPY fxj_ SCAN PRINT OUT =- OTHER (Specify)

F-ICOVERAGE BREAKDOWN Comments:

Procedure limited for the detection of axial flaws per the PDI issued PDQS document, CW/CCW scans were peifromed and areas of limitation were fu ___

examined _by-Lip1,enentai Eddy Current Technues.------.-

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Attachment A Docket No. 50-247 NL-08-053 Page 38 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 4 RV Nozzle Circ Weld RCC-22-1 (Datasheet)

REACTOR VESSEL WELD RESULTS

SUMMARY

PLANT NAME Indian Point Unit 2 WELD NO. RCC-22-1 COMPONENT Outlet Pipe @ 158" LIMITATIONS: NO E] YES [F] See Coverage Breakdown Sheet RESULTS NO. OF INDICATIONS N/A NI X STATUS N/A RI EXAM DOCUMENTATION INDICATION DOCUMENTATION

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-,- PARAGON ACQUISITION LOG IF] PARAGON HARD COPY L.] SCAN PRINT OUT =. OTHER (Specify)

=t- COVERAGE BREAKDOWN Comments.

Procedure limited for the detection of axial flaws per the PDI issued PDOS document. CW/CCW scans were perfromed and areas of limitation were full_

examined by supjlemental Eddy Current Techniques.

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(n Perpendicular 98.00 98.00 Parallel Ln 79.00 77.00 AVERAGE 88.50 87.50 Comments: Procedure limited for the detection of axial flaws per the PDI issued PDQS document. CW/CCW scans were perfromed and areas of limitation were fully 0 0

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Attachment A Docket No. 50-247 NL-08-053 Page 52 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 4 RV Nozzle Circ Weld RCC-23-1 (Datasheet)

REACTOR VESSEL WELD RESULTS

SUMMARY

PLANT NAME Indian Point Unit 2 WELD NO RCC-23-1 COMPONENT Outlet Pipe @ 338T LIMITATIONS. NO fj YES [X7 See Coverage Breakdown Sheet RESULTS NO. OF INDICATIONS N/A NI X STATUS N/A RI EXAM DOCUMENTATION INDICATION DOCUMENTATION

=XPARAGON ANALYSIS LOG ="-'ASSESSMENT SHEET 7]J PARAGON ACQUISITION LOG. .. L PARAGO.N. HARD CQPY 11 SCAN PRINT OUT = OTHER (Specify)

=COVERAGE BREAKDOWN Comments:

Procedure limited for the detection of axial flaws per the PDI issued PDQS document. CWtCCW scans were perfromed and areas of limitation were fully examined by supplemental Eddy Current Techniques.

Utility -

-eview Dae /--- -

Analyst , " -.... Date* "--4 ANII Review______

R.V. COVERAGE EST ATE BREAKDOWNS PLANT NAME Indian Point WesDyne WELD NO. RCC-23-1 0

-nL z0 International N N

0 01 CD n COMPONENT Outlet Pipe @ 3380 0

BEAM ANGLE BREAK DOWN EL BEAM DIRECTION 70 L _______________

0 CD Code MRP -o CD Perpendicular cn 99.00 99.00 Parallel 79.00 77.00 6 (A

=r mA 01 AVERAGE 89.00 88.00 La Comments: Procedure limited for the detection of axial flaws per the PDI issued PDQS document. CW/CCW scans were perfromed and areas of limitation were fully 0 0

examined by supplemental Eddy Current Techniques.

Limitation due to tapered area of the overlay 'n the ID- See sketch in Tab B -- I-

//I COMBINED AVERAGE 88.50 Analyst 4

/A -, / - Date Zýýc CD >D z

WesDyne International z0 C) 10 Year Reactor Pressure Vessel ISI N 01 N

Limited Examination Coverage Summary Table (D n

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RCC-23-1 Ax 206.81 2.1 99 N/A N/A N/A Lift-Off due to overlay N/A transition RCC-23-1 Circ 206".81 43.65 79 N/A N/A N/A Lift-Off due to overlay 100 -0 I__ _transition Comments: The Ultrasonic Examination is limited due to lift off in the transition slope of the overlay on the ID surface. (D CD U1 The perpendicular scans were limited in select areas around the circumference and not 360 degrees. The circ scans were limited 360 degree in the effected areas. Eddy Current examination was used to sun~nlement the limited areas and n~htained I(1fl/, eoen*

0 Examiner:. Date; ~~- (U CD C>

.7 CDZ Z 171 - 0'1 r90p C3

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Page 1 of 1 DATA ACQUISITION LOG # 338-SE Utility: I Entergy _Plant: Indian Point Unit: 2 outage: R17 Procedure No: PDI-ISI-254-SE Procedure Rev. No.: 2 Applicable Sensitivity Calibration Data Sheet No: DET-SE-01 UT Examiner Signature:;o Level: 1I" Date: 05-02-06 Data File Name Weld Noo, Index Scan Total # 'AVE' z

Gain Date Time Comments 0 Start Start of Signal Adj. N 0 Sweeps Amplitude (dB) &.2 (mmldd/yy)

N 0-CD)

WN338.SE-PRP-ON RCC-23-1 0. 113.75" 366 CH1-11 -3 DS 05-01-06 2227 CH2-9 WN338-SE-PAR-ON RCC-23-1 118.5': 0. 52 CH1-12 -3 DS 05-02-06 2022 CH2-10 -0 CH3-14 01 CH4-9 CD CD (D cn M) 01 Ene~yReview72$') 1L ae 1 0

Forni 1Date 5-' - 0 Foun 12.4 PDI-ISI-254-SE, Rev. 2 Date 5-F"14 .

D)CD>

S001 C)c T 9 CD 0 N\)

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noWEspynr Page 1 of ANALYSIS LOG # 338-SE Disk No: I N/A Utility: 1 Entergy Plant: I Indian Point Unit: I 2 Outage: I R17 Procedure No: I PDM-ISI-254-SE Procedure Rev. No.: 2 Weld No: I-RCC-23-1 I Weld Type: I SE I Exam. Surface- 1L0D Applicable Sensitivity Calibration Data Sheet No: ] DET-SE-01 IAcquisition Log No: i 338-SE I PARAGO N Anal. I Release: 6.1.6 UT Examiner Signature: I Level: (i II Date: i 05-02-06 z -A M

0 Data File Name N 0n UT" Beam Angle t NI RI RI Examiner ID I Date Channel Beam Direction Resolution / Comments / Limitations PI No. .. a CV.*.CUG, 0 M1 WN338-SE-PRP-ON 0

1 70 IN X DN 05-01-06 WN338-SE-PRP-ON 2 70 OUT X DN 05-01-06 WN338-SE-PRP-ON 3 0 NO WASTAGE DETECTED DN 05-01-06 0 WN338-SE-PRP-ON 4 0 NO WASTAGE DETECTED 0 CD DN 05-01-06 WN338-SE-PAR-ON 1 70 CCW CD CD X DN 05-02-06

-0 WN338-SE-PAR-ON 2 70 CW X ON 05-02-06 C WN338-SE-PAR-ON CD 3 70 CCW X DN 05-02-06 0 CI)

WN338-SE-PAR-ON 4 70 CW X 0.

ON 05-02-06 01 0

0 ANti Revie, :i*:- -Dat , -m- 00 Form 12.5 PDI-ISI.254-SE, Rev. 2 CDD (0 001 (0' 0 ILO 0 q0 - 9 3C

WES yfl ET Analysis Log: SE-338-1 Utility: Entergy Plant: Indian Point Unit: 2 I Outage: 2R17 Procedure No: WDI-STD-146 Weld No. RCC-23-1 WldT~ype; SAFE END I Procedure Rev. No.: 5 z0 n N M Applicable Sensitivity Calibration Data Sheet No: ET-1 Acquisition Log No: SE-338 N 11 ET Examiner Signature:: Level I Dat52-06 (D 0

CD Data ET ET Probe Scan NI RI RI Examiner ID 1 Date File Name Probe Direction Resolution I Comments I Limitations No. A~a rC.l 0 -0 Cx' WN338-SE-PAR-ON I CIRC X FS /'6-2-06 3 CD WN338-SE-PAR-ON 2 CIRC X FS / 5-2-06 0 CD WN-338-SE-PRP-ON 1 AXIAL X FS 15-2-06 WN-338-SE-PRP-ON 2 AXIAL X FS / 5-2-06 Cn 0

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Attachment A Docket No. 50-247 NL-08-053 Page 62 of 71 10 CFR 50.55a Relief Request RR-CRV-75 Attachment 4 RV Nozzle Circ Weld RCC-24-1 (Datasheet)

REACTOR VESSEL WELD RESULTS

SUMMARY

PLANT NAME Indian Point Unit 2 WELD NO. RCC-24-1 COMPONENT Outlet Pipe weld @ 22' LIMITATIONS: NO I YES __ see Coverage Breakdown Sheet RESULTS NO. OF INDICATIONS N/A NI X STATUS N/A RI EXAM DOCUMENTATION INDICATION DOCUMENTATION

-,PARAGONANALYSIS LOG Li ASSESSMENT SHEET i-- 1PARAGON ACQUISITION LOG LZPARAGONHARDCOPY

)K.'jSCAN PRINT OUT =i] OTHER (Specify)

-- COVERAGE BREAKDOWN Comments:

Procedure limited for the detection of axial flaws per the PDI issued PDOS document. CW/CCW scans were perforned and areas of limitation were fully examined by supplemental Eddy Current Technirqss.

Utility Revie - "

S)Anaiyst 4"' ' Date:

A~IReview'-

z n 0

N N

CD)

CD

-0 CD 0 CD (n

-M cn Ii 0

CD CD Z Z r9 C0 0 qo 9 CD C) K CP 4ý

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121.0 122.0 123.0 124.0 125.0 oC C >D

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N N 0 (D 01 CD 0"1 031 (D

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.n Profile Corre-tled Yh 7h (D Ln Current 114.36 0.26

-7

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Cn J0T C

R.V. COVERAGE EST -ATE BREAKDOWNS PLANT NAME Indian Point WesDyne WELD NO, RCC-24-1 z0 International N N

0

-n CD COMPONENT Outlet Pipe weld @ 220 0 BEAM ANGLE BREAK DOWN 2-L CD BEAM DIRECTIONj CD 70 L' 'P__,,,

0 CD

.0 C

CD Perpendicular 99.00 99.00 ....

Parallel 79.00 77.00 C,)

(A AVERAGE . 89.00 88.00 CD (D

Comments: Procedure limited for the detection of axial flaws per the PDI issued PDQS document. CW/CCW scans were perfromed and areas of limitation were fully

- 0 examined by supplemental Eddy Current Techniques. 0 Limitation due to tapered area of the overlay on the ID- See sketch in Tab B *

.I 00 COMBINED AVERAGE 88.50: Analyst / Date  %/ 4 'z (a

CD Z 9CD>

. / -/-

-40 N3 w 0

-40C1 -4>:

F-

-M,Wii-pyn-Page 1 of 1 DATA ACQUISITION LOG # 22-SE Utility: Entergy _ _Plant: Indian Point Unit: 2 O R Procedu. Po64-SE P edure Rev. No.: 2 Applicable Sensitivity CaliWration DataShtetNo.] DET-SE-01 C)

UT Examiner Signature: / Level: II Dt

• z 0 0

Data File Name Wel, N I.

index Scan Total # AVE' Gain I Date Time Comments N -n Start Start of Signal Adj. . N M1 Sweeps Amplitude (dB) . i (1) 0n C1 0

WN22-SE-PRP-ON RCC-24-1 0°  ! 113.75- 366 CHl-11 -3 DS 05-01-06 0009 0, 1 52 CH2-10 WN22-SE-PAR-ON RCC-24-1 118.5-- 0. 52 CHI-9 -3 JJJ 05-02-06 1631 i (D CH2-7 CH3-11 CD CH4-10 0

CD M,

-L DO, CD MD

/ //

0 0

Enter Date_ _

F_00 Form 12.4 PDf-JSi-254.-SE, Rev. 2 A!Rcii Date 0) . , =

W(0 c 0a, C I- N)

O3-*M I 001 W~ >

WEs X1 Page 1 of I ANALYSIS LOG # 22-SE

.......: .N.A .......

Utility; Entergy Pln;.ninPitUi: 2 Outage: IR17 Procedure o.: P.....IP-I2o5e , Rev. No.: 2 Weld No: I.RC 1 Weld-Type: SE Exam. Surface: ID 0 Applicable Sensitivity Calibration Data Sheet No:

• --;_. 1* 60DET-SE-01

...*I \

Acquisition Log No: 1

/ .e.. -- -o_--

22-SE j PARAGON Anal. Release:

Le Ili/ N Dat-e- 05-02-06-- 6.1.6 z0 C) n N

N Data File Name UT Beam Angle wI ,*/ RI 0"1 RI Examiner ID ! Date CD Channel Beam Direction Resolution I Comments / Limitations 0 0 0"1 WN22-SE-PRP-ON 1 ý70 iN X DN 05-01-06 0 03 WN22-SE-PRP-ON 2 70 OUT X DN 05-01-08 WN22-SE-PRP-ON C1 3 0 NO WASTAGE DETECTED DN 05-01-06 0,

r CD WN22-SE-PRP-ON 4 0 o NO WASTAGE DETECTED DN 05-01-06 CD CD 0

WN22-SE-PAR-ON 1 7i0CCW X DAM 05-02-06 -o WN22-SE-PAR-ON 2 70 CW X DAM 05-2-06 WN22-SE-PAR-ON 3 70 CCW X DAM 05-02-08 6 WN22-SE-PAR-ON 4 70CW X DAM 05-02-06 ;xi C~n

-4 01 CD 0

0 J~ J~-~

Enie~v Revew

/ ~. . Date 7~ CD >

Formn12,5 PDi-Ii-254-GE-Rev. 2 A~t~ R~wicw_____ _____ Date_____ CD Z .00 (00 "0

-4 0n 4>

msýw E S nl ET Analysis Log: SE-22-1 Utility: Entergy Plant: Indian Point Unit: 2 1 Outage: 2R17 JJ 0

Proceduwe No: WDI-STD-146 Procedure Rev. No,: 5 z 0 Weld No. RCC-24-1 Weld Type: SAFE END 0 N DO Applicable Sensitivity Calibration Data Sheet No: ET1 Acquisition Log No: SE.22 N C1 ET Examiner Signature: 1_ . Level .11.. . . . . . . . . Date: 5-3-06 0

- - ... .... ... _ CD 0"1

-7___________________

1j____________ (D Data ET ET Probe Scan Ni RI RI Examiner IDI Date Probe Direction 0.

File Name Resolution I Comments I Limitations No, 0)

WN22-SE-PAR-ON 1 RX JDF 15-306 (D zB WN22-SE-PAR-ON 21 CIRC X JDF / 5-3-06 3l JDF I-3-06 CD C.

WN-22-SE-PRP-ON 2 AXIAL X

(-

WN-22-SE-PRP-ON . 2 AXIAL X JDF 15-3-06 AL 01 M0 0

0 CD)

Entergy Lev-el III 2 '-a, -Date ___ At/ilI~ Date CD~ ZDZ

.7 (0 Ul 0C (

I-Ln 4C

.. 0 -.4 >

R.V. COVERAGE EST XTE BREAKDOWNS PLANT NAME Indian Point WesDyne WELD NO. RCC-24-1 z

International 0 N

N COMPONENT Outlet Pipe weld @ 220 CD, 0~

BEAM ANGLE BREAK DOWN CD BEAM DIRECTION

....... Code 70LI MRP J ..I " -

Q CD CD 0 -0 CD Perpendicular 99.00 99.00 (n IU3 Parallel 79.00 7-7.00 (D AVERAGE .... 89.00 818.00

-4 c0n Comments: Procedure limited for the detection of axial flaws per the PD1 issued PDQS document. CW/CCW scans were perfromed and areas of limitation were fully examined by supplemental Eddy Current Techniaues. 0 C

Limitation due to tapered area of the overlay on the ID- See sketch in Tab B COMBINED AVERAGE 88.50' I I "-4.-

Analyst Date _ /_/__ ___ ZoI-~0 (0 o CE) .0 "

0 P CD