ML20217G996

SER Accepting First 10-year Interval Inservice Insp Requests for Relief for Plant,Units 1 &

ML20217G996
Person / Time
Site:	Braidwood
Issue date:	10/14/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20217G984	List: ML20217G979 ML20217G996
References
NUDOCS 9910220036
Download: ML20217G996 (13)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE FIRST 10-YEAR INTERVAL INSERVICE INSPECTION REQUESTS FOR REllEF COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION. UNITS 1 AND 2 DOCKET NOS. STN 50-456 AND STN 50-457

1.0 INTRODUCTION

inservice inspection (ISI) of the American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel (B&PV) Code and applicable addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).10 CFR 50.55a(a)(3) states that attematives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed attematives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would resu!t in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including l supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice inspection of Nuclear Power Plant Components,' to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of the Section XI of the ASME Code incorporated by reference in the 10 CFR j 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and i modifications listed therein subject to Commission approval. l Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, )

information shall be submitted to the Commission in support of that determination and a request made for relief from tne ASME Code requirement. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose l l

9910220036 991014 PDR ADOCK 05000456 P PDR

requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration

. to the burden upon the licensee that could result if the requirements were imposed. The Code

of record for Braidwood Station, Units 1 and 2 (Braidwood), first 10-year ISI interval, which began July 29,1988, and October 17,1988, respectively, is the 1983 Edition of ASME Code,Section XI.

Additionally, pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), licensees that make a determination that they are unable to completely satisfy the requirements for the augmented reactor vessel shell weld examination specified in 10 CFR 50.55a(g)(6)(ii)(A) shall submit information to the Commission to support the determination and shall propose an attemative to the examination requirements that would provide an acceptable level of quality and safety. The licensee may use the proposed attemative when authorized by the Director of the Office of Nuclear Reactor Regulation.

3y letter dated March 5,1998, Commonwealth Edison Company (the licensee) submitted to the NRC its requests for relief to the Section XI requirements for piping and components pursuant to 10 CFR 50.55a(g)(5) and its attemative to the Rule required augmented reactor pressure vessel shell welds examination. Three relief requests (NR-35, NR-36, and NR-37) were requested for Unit 2. Relief Request NR-38 was requested for Units 1 and 2. The licensee requested relief from the ASME Code,Section XI requirements for volumetric examination of various reactor pressure vessel welds due to interferences which made performance of the examinations impractical. The licensee's proposed alternative to the augmented examination of reactor vessel shell weld 2RV-02-002 volumetric examination of ' essentially 100W of the subject weld in the reactor vessel is a best-effort examination resulting in limited examination coverage of the weld that proudes an acceptable level of quality and safety. The NRC staff reviewed and evaluated the licensee's proposed reliefs pursuant to 10 CFR 50.55a(g)(5) and 10 CFR 50.55a(g)(6)(ii)(A)(5) for Braidwood, Units 1 and 2.

2.0 EVALUATION The NRC staff, with technical assistance from its contractor (Idaho Engineering and Environmental Laboratory (INEEL)), has reviewed the information conceming ISI program requests for relief submitted for the first 10-year interval for Braidwood in a letter dated March 5, 1998. The staff adopts the evaluations and recommendations for granting relief or authorizing attematives contained in the Technical Letter Report (TLR) prepared by INEEL.

The staff finds that the relief requests as evaluated by this safety evaluation provide reasonable assurance of the pressure integrity of the piping and components. The staff has determined that authorizing attematives and granting reliefs pursuant to 10 CFR 50.55a(3)(i) and 10 CFR 50.55a(g)(6)(i) may be authorized and granted by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest. In making these determinations, the staff has considered the impracticality of performing the required inspection and testing and the burden on the licensee if the requirements were imposed.

.. 1 The granting of relief is based upon the fulfillment of any commitments made by the licensee in l It's basis for each relief request and the attematives proposed. Program changes involving new or revised relief request must be submitted to the NRC for review.

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3.0 CONCLUSION

The Braidwood ISI program requests ior relief from the Code requirements have been reviewed l by the staff with the assistance of its contractor, INEEL. The attachment provides INEEL's

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technical evaluation of these relief requests. The staff has reviewed the contractor's report and concurs with the evaluations and recommendations for granting relief or authorizing altematives. The authorizing of altematives or granting of relief is based upon the fulfillment of any commitments made by the licensee in its basis for each relief request and the altematives proposed.- The implementation of the ISI program and relief requests is subject to inspection by the NRC, t

Relief may be granted for NR-35, NR-37, and NR-38 pursuant to 10 CFR 50.55a(g)(6)(i) for the remainder of the current 10-year interval. In making this determination, the staff considered the t impracticality of performing the required inspections and the burden on the licensee if the Code

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l requirements were imposed. The alternative proposed in relief request NR-36 is authorized for the current interval pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) and 10 CFR 50.55a(3)(i) on the basis that it provides an acceptable level of quality and safety.

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Attachment:

Technical Letter Report Principal Contributor: G. Hatchett Date: October 14, 1999.

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TECHNICAL LETTER REPORT ON FIRST 10-YEAR INTERVAL INSERVICE INSPECTION REQUEST FOR RELIEF NOS. NR-35 THROUGH NR-38 l FOR I COMMONWEALTH EDISON COMPANY l BRAIDWOOD NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NUMBERS: 50-456 AND 50-457

1. INTRODUCTION By letter dated March 5,1998, the licensee, Commonwealth Edison Company, submitted Request for Relief Nos. NR-35 through NR-38, seeking relief from the requirements of the ASME Code,Section XI, for the Braidwood Nuclear Power Station (NPS), Units 1 and

2. These requests for relief are for the first 10-year inservice inspection (ISI) interval.

The Idaho National Engineering and Environmental Laboratory (INEEL) staff's evaluation ,

of the subject request for relief is in the following section.

2. EVALUATION  :

The information provided by Commonwealth Edison Company in support of the requests for relief from Code requirements has been evaluated and the bases for disposition are  ;

documented below. The Code of record for the Braidwood NPS, Units 1 and 2, first 10-year ISI intervals, which began July 29,1988, and October 17,1988, respectively, is the 1983 Edition through Summer 1983 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code.

2.1 Reauest for Relief NR-36 Examination Cateaorv B-A Item B1.10. Auamented Reactor Pressure Vessel Examination oer 10 CFR 50.55afa)(6)(ii)(A)

Reaulatorv Reauirement: 10 CFR 50.55a(g)(6)(li)(A)(2) requires that all licenses shall augment Reactor Pressure Vessel (RPV) examination by implementing once, as part of the inservice inspection interval in effect on Septeinber 8,1992, the examination requirements for the RPV shell welds specified in item B1.10 of Examination Category B-A, " Pressure Retaining Welds in Reactor Vessel", in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code. For the purpose of this augmented examination, " essentially 100%" as used in Table IWB-2500-1 means more than 90% of the examination volume of each weld, where the reduction in coverage is due to interference by another component, or part geometry.

Licensee's Proposed Altemative Examination: Pursuant to 10 CFR 50.55a(a)(3)(i) the licensee has proposed an attemative to the coverage requirements of the augmented RPV examination required by the regulations becausa essentially 100%

coverage could not be achieved for Weld 2RV-02-002. The licensee stated:

"The ultrasonic examination of the Braidwood Unit 2 RPV shell weld,2RV-02-002, was performed to the maximum extent practical. In conjunction with the partial

ultrasonic examination, a supplemental VT-1 of the RPV shell wcld was conducted from the interior of the RPV."

Pursuant to 10 CFR 50.55a(g)(ii)(A)(5), relief is requested from the requirement to examine more than 90% of the examination volume of the RPV circumferential shell weld,2RV-02-002, on the bcsis that the altemative to the examination requirements would provide an acceptable level of quality and safety.

Licensee's Basis for Proposed Altamative:

• 10 CFR 50.55 a(g)(6)(ii)(A)(1) revokes all relief requests with respect to volumetric examination coverage for RPV shell welds specified in item B1.10. Portions of a previously granted First Interval relief request, NR 9, addressed limited exam coverage on the Braidwood RPV shell welds.

" Augmented examination of the subject RPV shell weld was conducted on Braidwood Unit 2 during A2RO6 refuel outage (Fall 1997). During this exam at Braidwood Unit 2, physical obstructions and geometry prevented ultrasonic (UT) coverage in excess of 90% of the required volume. The examination of the Lower Shell Course-to-Dutchman weld,2RV-02-002, is restricted by six (6) core barrel locating lugs welded to the inner surface of the vessel approximately 2.5 inches above the weld centerline (See Attachment 1). These lugs obstruct the automated UT examination tool from examining the Code required volume of weld and base material under and below each lug in both the circumferential and perpendicular scan directions (156' total for all 6 lugs, See Attachment 2, 3, and 4). All weld metal and base material can be examined between the lugs (204* total length between all 6 lugs). The 6 lug interferences limit the examination aggregate volume coverage for the weld and adjacent base metal to approximately 81% of the Code required volume. Compliance with the applicable Code requirements may be accomplished by redesigning and modifying the RPV and/or the building structure surrounding the vessel (s). Access for manualinspections from the OD of the RPV is limited because of the close proximity of the building structure to the RPV shell ,

(See Attachment 1), j

• Examination of the Code required examination volume was completed to the maximum extent practical using attemate UT techniques qualified to the highest standards available. RPV examinations were conducted from the I.D. of the vessel. Access to allow examination from the O.D. shell side of these welds is  ;

restricted due to the structural concrete surrounding the vessel. The examination i techniques employed have been demonstrated and qualified to the Performance l demonstration initiative (PDl) Program which meets the intent of the rules of Appendix Vill of the ASME Code,Section XI,1990 Edition with 1993 Addenda. 1 These techniques were used in place of the currently required Section XI,1983 )

Edition with Summer 1983 Addenda, techniques (Reference Relief Request NR-  ;

I 29). Although the techniques have been qualified at PDI for single direction scanning, examinations were performed from two directions, when required (i.e.,

performed from both sides of the weld on the same surface where feasible).

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" Strict ASME Section ill quality controls were used when designing, fabricating, and j installing these RPV welds.' Preservice examinations to the fullest extent practical were performed on these welds. Preservice Inspection (PSI) relief request 2NR-9 was submitted to the staff and approved for these lug interferences. Comed has recently performed these ultrasonic examinations to the fullest extent practical, i.e.

81%, during the A2RO6 refuel outage and no unacceptable indications to the applicable Section XI standards were detected. The results of the examination provide further assurance that inservice flaws have not developed in the subject l weld, in addition to UT, visual examinations (VT-1 and VT-2) of the weld also verifies its integrity. Thus, the modification of the RPV and/or the building structures to increase examination volume coverage from 81% to more than 90% would incur unnecessary radiological exposure and significant engineering expenses.

Braidwood Station believes this course of action is a hardship without a compensating increase in the level of quality and safety."

Evaluation: To comply with the augmer ted reactor vessel examination requirements of 10 CFR 50.55a(g)(6)(ii)(A), licensees must volumetrically examine essentially 100% of each of the item B1.11 shell welds. In accordance with the regulations, essentially 100% is defined as greater than 90% of the examination volume of each weld.

At Braidwood NPS, Unit 2, the augmented coverage requirements cannot be met for the 2RV-02-002 circumferential shell weld due to physical restrictions that limit scan coverage. The geometric configuration of the core barrellocating lugs limit coverage to 81% of the required examination volume. To obtain an additional 9%

increase in coverage for the subject weld, reactor vessel design modifications would be required to allow access from the inside surface. Therefore, imposition of this requirement would result in a significant hardship for the licensee.

As a result of the augmented volumetric examination rule, licensees must make a reasonable effort to maximize examination coverage of their reactor vessels. In cases where examination coverage from the inside surface is inadequate, examination from the outside surface using manualinspection techniques may be an onfion. However, at Braidwood NPS, Unit 2, the building structure limits access to 79 outside surface. Therefore, the licensee is unable to enhance coverage by examining from the outside surface.

The licensee has examined a significant portion of the subject weld in accordance with the PDi techniques proposed in relief request NR-29. Relief request NR-29 was granted by the NRC staff in the SER dated 5/13/97. Examination of weld 2RV-02-002 was completed during the A2R06 refueling outage and no unacceptable indications were detected using the PDI techniques. Furthermore, i the volumetric examination is supplemented with visual examinations (VT-1 and VT-2) of the wold. Based on the volumetric examination coverage attained in combination with the visual examination, the INEEL staff concludes that any significant patterns of degradation, if present, would have been detected and that the examinations performed provide reasonable assurance of the continued  !

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i structural integrity of the subject lower head to shell weld. Therefore, based on the amount of coverage (81%) obtained, and the inaccessibility of the weld from the outside surface, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

2.2 Reauest for Relief NR-35. Examination Cateaory B-A. Item B1.11. Pressure Retainina Welds in Reactor Vessel Code Reauirement: Subsection IWB, Table IWB-2500-1, Examination Category B-A, item Number 81.11 reauires essentially 100% volumetric examination of the J RPV Circumferential Shell Welds as detailed in Figure IWB-2500-1.

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l Licensee's Code Relief Reauest: Pursuant to 10 CFR 50.55a(g)(5)(iii) , relief is i requested from examining 100% of the Code-required volume of reactor vessel Weld 2RV-02-002. I Licensee's Basis for Relief Peauest (as statedt "10 CFR 50.55 a(g)(6)(ii)(A)(1) revokes all relief requests with respect to volumetric examination coverage for welds specified in item B1.10. Portions of a previously granted relief request, NR-9, addressed limited exam coverage on the Braidwood RPV shell welds.

" Examination of the subject RPV shell weld was conducted on Braidwood Unit 2 during A2 ROB refuel outage (Fall 1997). During this exam at Braidwood Unit 2, physical obstructions and geometry prevented ultrasonic (UT) coverage in excess of 90% of the required volume. The examination of the Lower Shell Course-to-Dutchman weld,2RV-02-002, is restricted by six (6) core barrel locating lugs welded to the inner surface of the vessel approximately 2.5 inches above the weld centerline (See Attachment 1)' . These lugs obstruct the automated UT examination tool from examining the Code required volume of the weld and bat.e material unde and below each lug in both the circumferential and perpendicult.r scan directions (156' total for all 6 lugs, See Attachment 2, 3 and 4). All weld metal and base material can be examined between the lugs (204* total length between all 6 lugs). The 6 lug interferences limit the examination aggregate volume coverage obtained for the weld and adjacent base metal to approximately 81% of the Code required volume.

" Compliance with the applicable Code requirements may be accomplished by redesigning and modifying the RPV and/or the building structure surrounding the vessel (s). Access for manual inspections from the CD of the RPV is limited because of the close proximity of the building structure to the RPV shell (See Attachment 1).

1 Tables, Figures and attachments fumished with the licensee's submittal are not included in this report.'

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" Examination of the Code required examination volume was completed to the maximum extent practical using attemate UT techniques qualified to the highest standard available. RPV examinations were conducted from the I.D. of the vessel.

Access to allow examination from the O.D. (shell side) of these welds is restricted due to the structural concrete surrounding the vessel. The examination techniques employed have been demonstrated and qualified to the Performance l Demonstration Initiative (PDI) Program which meets the intent of the rules of Appendix Vill of the ASME Code,Section XI,1992 Edition with 1993 Addenda.

These techniques were used in place of the currently required Section XI,1983 Edition with Summer 1983 Addenda, techniques (Reference Relief Request NR-29). Although the techniques have been qualified at PDI for single direction scanning, examinations were performed from two directions, when required (i.e.,

performed from both sides of the weld on u J same surface, where feasible). Strict ASME Section 111 quality controls were used when designing, fabricating, and installing these RPV welds. Preservice (PSI) examinations to the fullest extent j practical were performed on these welds. PSI relief request 2NR-9 was submitted l to the Staff and approved for these lug interferences. Comed has recently  ;

performed these ultrasonic examinations to the fullest extent practical, i.e. 81%, '

during the A2RO6 refuel outage and no unacceptable indications to applicable Section XI standards were detected. The results of the examination provide further assurance that unallowable inservice flaws have not developed in the subject weld.

In addition to UT, visual examinations (VT-1 and VT-2) of the weld also verifies its integrity. Thus, the modification of the RPV and/or the building structures to increase 3xamination volume coverage from 81% to essentially 100% would incur .

unnecessary radiological exposure and significant engineering expenses, l "Braidwood Station believes this course of action is a hardship without a ,

compensating increase in the level of quality and safety." i Licensee's Proposed Attemative Examination:

"The ultrasonic examination of the Braidwood Unit 2 RPV shell weld,2RV-02-002, was performed to the maximum extent practical. In conjunction with the partial ultrasonic examination, a supplemental VT-1 of the RPV shell weld was conducted from the interior of the RPV using underwater camera equipment."

Evaluation: The Code requires that the subject reactor pressure vessel circumferential shell weld be 100% volumetrically examined during the inspection interval. Examination coverage from inside the reactor vessel was limited to 81% 1 due to the geometric configuration of the core barrel locating lugs. Manual inspection from outside the reactor vessel is restricted due to the close proximity of the building structure to the reactor vessel shell. Based on the information provided in this request for relief, the INEEL has determined that it is impractical to examine the subject welds to the extent required by the Code. For complete examination coverage, redesign and modification of the reactor vessel and/or building structure would be necessary. ,c.psition of this requirement would cause a considerable burden on the licensee.

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The licensee proposes to perform the volumetric examinations to the maximum extent practical on the subject welds in accordance with the PDI techniques

, proposed in relief request NR 29. Relief request NR-29 was granted by the NRC I

staff in an SER dated 5/13/97. Examination of weld 2RV-02-002 was completed during the A2R06 refueling outage and no unacceptable indications were detected using the PDI techniques. Based on the significant portion of the volumetric examinations completed, it is reasonable to conclude that a pattern of degradation, l

if present, would have been detected. As a result, reasonable assurance of continued structural integrity has been provided. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).

l 2.3 Reauest for Relief NR-37. Unit 2. Examination Cateoorv B-D. Item B3.90. Full Penetration Welds of Nonles in Vessels l l Code Reauirement: Examination Category B-D, item Number B3.90 requires l l essentially 100% volumetric examination of reactor pressure vessel nozzle welds, as defined by Figure IWB-2500-7.

Licensee's Code Relief Reauest: In accordance with 10 CFR 50.55a(g)(5)(iii), the  !

licensee requested relief from the Code-required volumetric examination for the i nozzle welds listed below.

1 WELD ' ITEM ' DESCRIPTION COVERAGE LIMITATION I 2RV-01-006 B3.90 Hot Leg Nozzle 81 % integral Extension 2RV-01-009 B3.90 Hot Leg Nozzle 81 % integral Extension 2RV-01-010 B3.90 Hot Leg Nozzle 81 % integral Extension 2RV-01-013 B3.90 Hot Leg Nozzle 81 % integral Extension

' Licensee's Basis for Relief Reauest (as statedt "All RPV welds are examined using remotely operated underwater volumetric inspection techniques. Underwater volumetric inspection techniques are utilized to meet ALARA concems due to the high radiation levels in these areas. The outlet (Hot Leg) nozzles are constructed with an integral extension on the I.D. surface which mates with the intemal core barrel. The extension provides a flow path for reactor coolant from the core into the hot leg nozzles. The integral extensions j partially obstructs the circumferential scan for reflectors transverse to the weld j (Reference Attachment 1). The integral extension, that confines the movement of j the transducer package, along with the curvature of the RPV shell combine to limit  !

full Code volume coverage when scanning in the direction parallel to the weld l (Reference Attachment 2). This configuration limits the examination aggregate l volume coverage obtained for each weld and adjacent base metal to approximately l 81% instead of the Code required essentially 100%, examination coverage."  ;

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" Compliance with the applicable Code requirements may be accomplished by redesigning and modifying the ID of the Hot Leg nozzles and/or the building structure surrounding the RPV at the nozzles' elevation. Braidwood Unit 2 RPV was designed with a RPV shield wall (Reference Attachment 3 and 4). This wall impedes access to the OD of the RPV shell for insulation removal, surface preparation and ultrasonic inspection. Modifying the nozzle ID surface would incur extensive radiation exposure to station personnel and could be detrimental to the component. When designing, fabricating and installing these welds, strict ASME Section lll quality controls ano procedures were used that minimized the introduction of fabrication defects. Additionally, the periodic VT-2 examinations in accordance with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-P and applicable Reactor Coolant system, monitoring requirements specified in the Technical Specifications will provide reasonable assurance of continued structuralintegrity of the Reactor vessel. Comed has recently performed these volumetric examinations to the fullest extent practical, i.e.

81%, during the A2R06 refuel outage and no recordable indications (NRI) were detected. The NRI results of the examination provide further assurance that unacceptable inservice flaws have not developed in the subject welds. Thus, the modification of the nozzles and/or the building structure to increase examination volume coverage from 81%, to essentially 100%, would incur unnecessary radiological exposure and significant engineering costs without a compensating increase in the level of quality and safety."

Licensee's Proposed Alternative Examination:

"The Reactor vessel outlet (Hot Leg) nozzle welds will be examined to the fullest extent practical using the available underwater volumetric inspection techniques."

Evaluation: The Code requires volumetric examination of essentially 100% of the RPV nozzle to vessel welds. The licensee performs the required volumetric l examinations from the inside of the nozzle since access to the weld from the outside is prevent?d by the RPV shield wall and the reactor vessel supports.

Complete examination of these welds from the inside of the nozzles is not practical due to the interference caused by integral extensions. Significant modifications of the nozzles and/or the building structure would be necessary to enable volumetric examination of essentially 100% of the Code required volume. Imposition of this requirement would place a substantial burden on the licensee.

The licensee completed a significant portion (81%) of the required examination volume during the recent A2R06 refueling outage and no unacceptable indications were detected. These welds also receive visual examinations (VT-2) during each refueling outage. Any patterns of degradation would have been detected by the visual and volumetric examinations and reasonable assurance of structuralintegrity and operational readiness of the subject nozzle-to-vessel welds has been provided.

Therefore, based on the impracticality of meeting the Code coverage requirements for the subject welds, and the reasonable assurance provided by the examinations

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that were completed, it is recommended that relief be granted pursuant to 10CFR50.55a(g)(6)(i).

2.4 Reauest for Relief NR-38. Examination Cetecorv B-A. Item B1.21. Circumferential Head Welds Code Reauirement: Subsection lWB, Table IWB-2500-1, Examination Category B-A, item Number B1.21 requires essentially 100% volumetric examination of the accessible length of the Reactor Pressure Vessel (RPV) Lower Head Circumferential Weld as detailed in Figure IWB-2500-3.

Licensee's Code Relief Reauest: Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code-required 100% volumetric examination coverage for the welds listed below.

WELD ITEM DESCRIPTION COVERAGE LIMITATION 1RV-02-001 B1.21 Unit 1, Lower 86 % Instrument Head Nonle Circumferential Penetrations Weld 2RV-02-001 B1.21 Unit 2, Lower. 86 % Instrument Head Nonle Circumferential Penetrations Weld Licensee's Basis for Relief Reauest:

" Examination of the subject RPV Lower Head Circumferential weld was conducted 1 on Braidwood Unit 1 during A1RO6 refuel outage (Spring 1997) and on Braidwood Unit 2 during A2R06 refuel outage (Fall 1997) During these exams at Braidwood Units 1 and 2, physical obstructions and geometry prevented ultrasonic (UT) coverage in excess of 90% of the accessible length of weld volume. The examination of the Lower Head Circumferential welds,1RV-02-001 and 2RV-02-001, are restricted by instrumentation nonle penetrations (See Attachment 2). These instrumentation nonle penetrations obstruct the automated UT examination tool from examining the accessible length of Code weld and base material volume in both the circumferential and perpendicular scan directions. All weld metal and base material can be examined between the instrumentation nonle penetrations. The instrumentation nonle penetrations interferences limit the examination aggregate volume coverage obtained for the subject welds and adjacent base metal to approximately 86% of the Code required volume.

" Compliance with the applicable Code requirements may be accomplished by redesigning and modifying the RPV and/or the building structure surrounding the a-

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vessel (s). Access for manual inspections from the OD of the RPV is limited because of the close proximity of the building structure to the RPV shell(See Attachment 1).

" Examination of the Code required examination volume was completed to the maximum extent practical using altomate UT techniques qualified to the highest standard available. RPV examinations were conducted from the I.D. of the vessel.

Access to allow examination from the O.D. of these welds is restricted due to the structural concrete surrounding the vessel. The examination techniques employed have been demonstrated and qualified to the Performance Demonstration Initiative (PDl) Program, which meets the intent of the rules of Appendix Vill of the ASME Code,1992 Edition with 1993 Addenda. These techniques were used in place of the currently required Section XI,1983 Edition with Summer 1983 Addenda, techniques (Reference Relief Request NR-29). Although the techniques have been qualifeed at PDI for single direction scanning, examinations were performed from two directions, when required (i.e., performed from both sides of the weld on the same surface, where feasible).

" Strict ASME Section til controls were used when designing, fabricating, and installing these RPV welds. Preservice (PSI) examinations to the fullest extent practical were performed on these welds. PSI relief requests 1(2)NR-9 were submitted to the Staff and approved for these tag intourences. Comed has recently Performed these ultrasonic examirations to the fullest extent practical, i.e.

86%, during the A1R06 and A2R06 refuel outages and no unacceptable indications to applicable Section XI standards were detected. The results of these examinations provide further assurance that unallowable inservice flaws have not developed in the subject welds. In addi%n to UT, visual examinations (VT-2) of the weld also verifies its integrity. Thus, me modification of the RPV and/or the building structures to increase examination volume coverage from 86% to essentially 100%; would incur unnecessary radiological exposure and significant engineering expenses. Braidwood Station believes this course of action is a hardship without a compensating increabe in the level of quality and safety."

Licensee's Proposed Alterrntive Examination. <

"The ultrasonic examination of the Braidwood Units 1 and 2 Lower Head l Circumferential Welds,1RV-02-001 and 2RV-02-001, will be performed to the I maximum extent practical." i Evaluation: The Code requires volumetric examination of essentially 100% of the RPV head welds. However, complete examination of these welds from the inside of the RPV is not practical due to interference from the instrumentation nozzle penetrations. The required examinations cannot not be performed from outside the RPV because access is limited due to the close proximity of the building structure to the RPV shell. Significant modifications of the vessel and/or the building structure would be necessary to enable volumetric examination of essentially 100%

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I V e i j D of the Code required volume. Imposition of this requirement would impose an undue burden on the licensee.

The licensee completed a significant portion (86%) of the required examination ,

volume of welds 1RV-02-001 and 2RV-02-001 during the recent A1R06 and A2R06 i refueling outages in accordance with the PDI techniques proposed in relief request NR-29. Relief request NR 29 was granted by the NRC staffin the SER dated 5/13/97. No unacceptable indications were detected during recent volumetric examinations. In addition these welds have also received visual examinations (VT-

2) during the refueling outages. Therefore, any existing patterns of degradation wsuld have been detected by the visual and volumetric examinations that were completed and reasonable assurance of structuralintegrity and operational readiness has been provided.

Based on the impracticality of meeting the Code coverage requirements for the subject welds, and the reasonable assurance provided by the examinations that were completed, it is recommended that relief be granted pursuant to 10CFR50.55a(g)(6)(i).

3. CONCLUSION l The INEEL staff has reviewed the licensee's submittal and concludes that for Request for Relief No. NR-36, that the significant coverage (81% on one weld, and 100% on all other f

shell welds) obtained provides an acceptable level of quality and safety. Therefore, it is recommended that the proposed attematives be authorized pursuant to 10 CFR 50.55a(a)(3)(i). For Request for Relief Nos. NR 35, NR-37 and NR-38, the INEEL staff concludes that the Code requirements are impractical to perform to the extent required.

Therefore, it is recommended that relief be granted pursuant to 10CFR50.55a(g)(6)(i).

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