RA-12-097, Submittal of Relief Requests for the Fifth Ten-Year Inservice Inspection (Lsi) Interval

Submittal of Relief Requests for the Fifth Ten-Year Inservice Inspection (Lsi) Interval

ML12243A287
Person / Time
Site:	Oyster Creek
Issue date:	08/28/2012
From:	Jesse M Exelon Generation Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-12-097
Download: ML12243A287 (48)

Text

10 CFR 50.55a RA-12-097 August 28, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

Submittal of Relief Requests for the Fifth Ten-Year Inservice Inspection (lSI)

Interval Attached for your review are five (5) relief requests associated with the fifth ten-year Inservice Inspection (lSI) interval for Oyster Creek Nuclear Generating Station (OCNGS). The fifth ten-year lSI interval of the OCNGS lSI program will comply with the 2007 Edition through 2008 Addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI. The fifth ten-year lSI interval will begin on January 15, 2013. We request your approval by August 28, 2013.

There are no regulatory commitments in this letter.

If you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Respectfully, Attachments: 1) Relief Request 15R-01

2) Relief Request 15R-02

3) Relief Request 15R-05

4) Relief Request 15R-06

5) Relief Request 15R-07 cc: Regional Administrator, Region I, USNRC USNRC Senior Resident Inspector, OCNGS USNRC Senior Project Manager, OCNGS

Attachment 1 Relief Request I5R-Ol

10 CFR 50.55a RELIEF REQUEST 15R"01 Revision 0 (Page 1 of 32)

Request for Relief for Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection In Accordance with 10 CFR 50.55a(a)(3)(i)

1. ASME Code Component(s) Affected:

Code Class: 1

Reference:

IWB-2500 Table IWB-2500-1 Examination Category: B-N-l and B-N-2 Item Number: B13.10, B13.20, B13.30, and B13.40

Description:

Use of BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection Component Numbers: Vessel Interior. Interior Attachments within Beltline Region, Interior Attachments beyond Beltline Region, and Core Support Structure

2. Applicable Code Edition and Addenda:

The Oyster Creek Nuclear Generating Station Inservice Inspection (lSI) program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirements:

ASME Section XI requires the examination of components within the reactor pressure vessel. These examinations are included in Table IWB-2500-1, Categories B-N-l and B-N-2 and identified with the following item numbers:

B13.10 Examine accessible areas of the reactor vessel interior each period by the VT-3 visual examination method (B-N-l).

B13.20 Examine interior attachment welds within the beltline region each interval by the VT-l visual examination method (B-N-2).

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B13.30 Examine interior attachment welds beyond the beltline region each interval by the VT-3 visual examination method (B-N-2).

B13.40 Examine surfaces of the welded core support structure each interval by the VT-3 visual examination method.

These examinations are performed to assess the structural integrity of components within the boiling water reactor pressure vessel.

4. Reason for Request

Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested for the proposed alternative to the Code requirements provided above on the basis that the use of the BWRVIP guidelines discussed below will provide an acceptable level of quality and safety.

The BWRVlP Inspection and Evaluation (I&E) guidelines have recommended aggressive specific inspection by BWR operators to completely identify material condition issues with BWR components. A wealth of inspection data has been gathered during these inspections across the BWR industry. I&E guidelines focus on specific and susceptible components, specify appropriate inspection methods capable of identifying real anticipated degradation mechanisms, and require re-examination at conservative intervals. In contrast, the code inspection requirements were prepared before the BWRVlP initiative and have not evolved with BWR inspection experience.

Use of this proposed alternative will maintain an adequate level of quality and safety and avoid unnecessary inspections.

5. Proposed Alternative and Basis for Use:

In lieu of the requirements of ASME Section XI, the proposed alternative is detailed in Table 1 for Examination Category B-N-l and B-N-2.

Oyster Creek Nuclear Generating Station will satisfy the Examination Category B-N-l and B-N-2 requirements as described in Table 1 in accordance with BWRVlP guideline requirements. This relief request proposes to utilize the associated BWRVIP guidelines in lieu of the associated Code requirements including but not limited to exam method, volume, frequency, training, successive and additional examinations, flaw evaluations, and reporting.

Not all the components addressed by these guidelines are code components. The following guidelines are applicable to this Relief Request:

- BWRVlP-03, "BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines"

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BWRVIP-18, Revision 1, "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines"

- BWRVIP-26-A, "BWR Top Guide Inspection and Flaw Evaluation Guidelines"

- BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate ~P Inspection and Flaw Evaluation Guidelines" BWRVIP-38, "BWR Shroud Support Inspection and Flaw Evaluation Guidelines" BWRVIP-47-A, "BWR Lower Plenum Inspection and Flaw Evaluation Guidelines"

- BWRVIP-48-A, "Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines" BWRVIP-76, Revision I, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines"

- BWRVIP-94, Revision 2, "BWR Vessel and Internals Project Program Implementation Guide"

- BWRVIP-183, "BWR Vessel and Internals Project, Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines" Table 1 compares present ASME Examination Category B-N-I and B-N-2 requirements with the above current BWRVIP guideline requirements, as applicable, to the Oyster Creek Nuclear Generating Station BWRJ2 unit.

Inspection services, by an Authorized Inspection Agency, will be applied to the proposed alternative actions of this relief request.

BWRs now examine reactor internals in accordance with BWRVIP guidelines. These guidelines have been written to address the safety significant vessel internal components and to examine and evaluate the examination results for these components using appropriate methods and reexamination frequencies. The BWRVIP has established a reporting protocol for examination results and deviations. The NRC has agreed with the BWRVIP approach in principal and is expected to issue Safety Evaluations for many of these guidelines (see References 1 - 9 below). Therefore, use of these guidelines, as an alternative to the subject Code requirements, provides an acceptable level of quality and safety and will not adversely impact the health and safety of the public.

As additional justification, Appendix 1 ("Comparison of Code Examination Requirements to BWRVIP Examination Requirements") provides specific examples which compare the inspection requirements of ASME Code Item Numbers BI3.IO, B13.20, B13.30, and B13.40 in Table IWB-2500-1, to the inspection requirements in the BWRVIP documents. Specific BWRVIP documents are provided as examples. This comparison also includes a discussion of the inspection methods. These comparisons demonstrate that use of these guidelines, as an alternative to the subject Code requirements, provides an acceptable level of quality and safety and will not adversely impact the health and safety of the public.

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Appendix 2 contains the "Reactor Vessel Inspection History" for Oyster Creek Nuclear Generating Station. This summary provides, on a component-by-component basis, the inspection methods utilized, the inspection dates, and the results of the inspections. This table also contains the identified corrective actions. The information provided reflects the compilation of the BWRVIP outage reports. Corrective actions and inspections performed prior to the BWRVIP were implemented to the requirements of ASME Section XI, as applicable.

The reactor vessel internals inspection program at Oyster Creek Nuclear Generating Station has been developed and implemented to satisfy the requirements of BWRVIP-94.

It is recognized that the BWRVIP executive committee periodically revises the BWRVIP guidelines to include enhancements in inspection techniques and flaw evaluation methodologies. Where the revised version of a BWRVIP inspection guideline continues to also meet the requirements of the version of the BWRVIP inspection guideline that forms the safety basis for the NRC-authorized proposed alternative to the requirements of 10 CFR 50.55a, it may be implemented. Otherwise, the revised guidelines will only be implemented after NRC approval of the revised BWRVIP guidelines or a plant-specific request for relief has been approved.

6. Duration of Proposed Alternative:

Relief is requested for the fifth ten-year lSI interval for Oyster Creek Nuclear Generating Station. The fifth ten-year lSI interval will begin on January 15,2013.

7. Precedents:

Similar relief requests have been approved for:

Vermont Yankee Nuclear Power Station as discussed in Reference 10. The Vermont Yankee Nuclear Power Station Relief Request for BWRVIP was authorized in the NRC SER dated September 19,2005.

Exelon/AmerGen BWR fleet as discussed in Reference 11. The Exelon/AmerGen Fleetwide Relief Request for BWRVIP was authorized in the NRC SER dated April 30, 2008.

Perry Nuclear Power Plant, Unit No.1 as discussed in Reference 12. Perry Nuclear Power Plant, Unit No.1 was authorized in the NRC SER dated January 31,2012.

Fermi 2 as discussed in Reference 13. Fermi 2 was authorized in the NRC SER dated February 17,2012.

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8.

References:

1. Letter from NRC to BWRVIP, "Final Safety Evaluation for Electric Power Research Institute Boiling Water Reactor Vessel and Internals Project Technical Report 1016568, 'BWRVIP-18, Revision 1: BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines (TAC No. ME2189)' ," dated January 30,2012 (ML113620684)

2. Letter from NRC to BWRVIP, "NRC Approval Letter of BWRVIP-26-A, 'BWR Vessel and Internals Project Boiling Water Reactor Top Guide Inspection and Flaw Evaluation Guidelines' ," dated September 9, 2005

3. Letter from NRC to BWRVIP, "Propriety Version of NRC Staff Review of BWRVIP-27-A, 'BWR Standby Liquid Control System/Core Plate LlP Inspection and Flaw Evaluation Guidelines' ," dated June 10,2004

4. Letter from NRC to BWRVIP, "Final Safety Evaluation of the 'BWR Vessel and Internals Project, BWR Shroud Support Inspection and Flaw Evaluation Guidelines (BWRVIP-38): EPRI Report TR-108823 (TAC No. M99638)," dated July 24, 2000

5. Letter from NRC to BWRVIP, "NRC Approval Letter of BWRVIP-47-A, 'BWR Vessel and Internals Project Boiling Water Reactor Lower Plenum Inspection and Flaw Evaluation Guidelines' ," dated September 9, 2005

6. Letter from NRC to BWRVIP, "NRC Approval Letter of BWRVIP-48-A, 'BWR Vessel and Internals Project Vessel ID Attachment Weld Inspection and Flaw Evaluation Guideline' ," dated July 25,2005

7. "BWRVIP-76NP, Revision 1: BWR Vessel and Internals Project BWR Core Shroud Inspection and Flaw Evaluation Guidelines," dated May 2011 (ML11195A182)

8. Letter from Chairman, BWR Vessel and Internals Project to NRC, "Project No. 704 - BWRVIP Program Implementation Guide (BWRVIP-94NP, Revision 2)," dated September 22, 2011 (MLI1271A058)

9. "BWRVIP-183NP: BWR Vessel and Internals Project Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines," dated December 7,2007 (ML080220433)

10. Letter from NRC to Entergy Nuclear Operations, Inc., "Safety Evaluation of Relief Request RI-Ol, Vermont Yankee Nuclear Power Station (TAC No.

MC0690)," dated September 19,2005 (ML052370244)

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11. Letter from NRC to Exelon/AmerGen, "Clinton Power Station, Unit 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Limerick Generating Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; Peach Bottom Atomic Power Station, Units 2 and 3; and Quad Cities Nuclear Power Station, Units 1 and 2 - Relief Request to Use Boiling Water Reactor Vessel and Internals Project Guidelines in Lieu of Specific ASME Code Requirements (TAC Nos. MD5352 through MD5363)," dated April 30, 2008 (ML080980311 )

1 Letter from NRC to FirstEnergy Nuclear Operating Company (Perry Nuclear Power Plant, Unit No.1), "Perry Nuclear Power Plant, Unit No. I, RE: Safety Evaluation in Support of 10 CFR 50.55a Requests for the Third 10-Year In-Service Inspection Interval (TAC Nos. ME5373, ME5376, ME5377, ME5379, and ME5380)," dated January 31, 2012 (MLI20180372)

13. Letter from NRC to Detroit Edison Company (Fermi 2), "Fermi 2 - Evaluation of Applicable to-Year Interval Inservice Inspection Relief Request - Use of Boiling Water Reactor Vessel and Internals Project (BWRVIP) Guidelines in Lieu of Specific ASME Code Requirements (TAC NO. ME6765)," dated Febrnary 17,2012 (ML120370286)

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TABLE 1 Comparison of ASME Examination Category B-N-l and B-N-2 Requirements With BWRVIP Guidance Requirements(l)

ASMEItem Component ASMEExam ASME ASME Authorized BWRVIP BWRVIP BWRVIP

~ ~

Number, Scope Exam Alternative Exam Table IWB- Scope 2500-1 B13.10 Reactor Vessel Interior Accessible VT-3 Each BWRVIP Overview examinations of during Areas period Rl, 26-A, BWRVIP examinations Code VT-3 visual 38, 47-A,48-A, inspection requirements.

and 76-RI B13.20 Interior Attachment Accessible VT-I Each BWRVIP-48-A, Bracket VT-I Each [0-:

Within Beltline Region Welds IO-year Table 3-2 Attachment Interval

- Lower Surveillance Interval Specimen Holder Brackets B13.30 Interior Attachments Accessible VT-3 Each Beyond Beltline - Welds IO-year Guide Rod Brackets Interval BWRVIP-48-A, Bracket VT-3 Each [0-:

Table 3-2 Attachment Interval Steam Dryer Support BWRVIP-48-A, Bracket EVT-I Each I0-year Brackets Table 3-2 Attachment Interval Feedwater Sparger BWRVIP-48-A, Bracket EVT-I Each 10-:

Brackets Table 3-2 Attachment Interval Upper Surveillance BWRVIP-48-A, Bracket VT-3 Each 10-:

Specimen Holder Table 3-2 Attachment Interval Brackets Shroud Support (Weld BWRVIP-38, Weld H9(2) EVT-I or UT Maximum of6 H9) 3.1.3.2, years for one-sided Figure 3-5 EVT-I, Maximum of 10 years for UT

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TABLE 1 Comparison of ASME Examination Category B-N-I and B-N-2 Requirements With BWRVIP Guidance Requirements(l)

ASMEItem BWRVIP Number, ASMEExam ASME ASME Authorized BWRVIP BWRVIP Component Exam ~

Tahle IWB- $cope Exam Frequency Alternative Exam Scope 2500-1 B13.40 Welded Core Support Accessible VT-3 Each BWRVIP-38, Shroud EVT-l or UT Based on as found Structure - Shroud Surfaces IO-year 3.1 Support to a Support Interval Figure 3-5 maximum 6 years ft)r one-sided EVT-I, 10 years for UT where accessible Shroud Vertical Welds BWRVIP Vertical and EVT-I or UT Maximum 6 years RI, Ring for one-sided 3.3, Segment EVT-l, 10 years for Figure 3-1 Welds as UT applicable Shroud Repairs(3) BWRVIP Tie-Rod VT-3 Per designer Rl, Repair recommendations Section 3.5 per BWRVIP RI NOTES:

1) This Table provides only an overview of the requirements. For more details, refer to ASME Section XI, Table IWB-2500-1, and the aPIPH1,priate BWRVIP document.

2) In accordance with Appendix A of BWRVIP-38, a site specific evaluation will determine the minimum required weld length to be examined.

3) Shroud repairs are currently installed on the Oyster Creek Nuclear Generating Station.

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Appendix 1 Comparison of Code Examination Requirements to BWRVIP Examination Requirements The following discussion provides a comparison of the examination requirements provided in ASME Code Item Numbers B13.10, BI3.20, B13.30, and B13.40 in Table IWB-2500-1, to the examination requirements in the BWRVIP guidelines. Specific BWRVIP guidelines are provided as examples for comparisons. This comparison also includes a discussion of the examination methods.

1. Code Requirement - B13.10 - Reactor Vessel Interior Accessible Areas (B-N-l)

The ASME Section Xl Code requires a VT-3 visual examination of reactor vessel accessible areas, which are defined as the spaces above and below the core made accessible during normal refueling outages. The frequency of these examinations is specified as the first refueling outage, and at intervals of approximately 3 years, during the first inspection interval, and each period during each successive ten-year lSI inspection. Typically, these examinations are performed every other refueling outage of the inspection intervaL This examination requirement is a non-specific requirement that is a departure from the traditional ASME Section Xl examinations of welds and surfaces.

As such, this requirement has been interpreted and satisfied differently across the domestic fleet. The purpose of the examination is to identify relevant conditions such as distortion or displacement of parts; loose, missing, or fractured fasteners; foreign material, corrosion, erosion, or accumulation of corrosion products; wear; and structural degradation.

Portions of the various examinations required by the applicable BWRVlP guidelines require access to accessible areas of the reactor vessel during each refueling outage.

Examination of core spray piping and spargers (BWRVIP-18-Rl), top guide (BWRVIP-26-A), interior attachments (BWRVIP-48-A), core shroud welds (BWRVIP-76-Rl), and shroud support (BWRVIP-38) provides such access. Locating and examining specific welds and components within the reactor vessel areas above, below (if accessible), and surrounding the core (annulus area) entails access by remote camera systems that essentially perform equivalent VT-3 visual examination of these areas or spaces as the specific weld or component examinations are performed. This provides an equivalent method of visual examination on a more frequent basis than that required by the ASME Section Xl Code. Evidence of wear, structural degradation, loose, missing, or displaced parts, foreign materials, and corrosion product buildup can be, and has been observed during the course of implementing these BWRVIP examination requirements.

Therefore, the specified BWRVlP Guideline requirements meet or exceed the subject Code requirements for examination method and frequency of the interior of the reactor

10 CFR 50.55a RELIEF REQUEST I5R-01 Revision 0 (Page lO of 32) vessel. Accordingly, these BWRVIP examination requirements provide au acceptable level of quality and safety as compared to the subject Code requirements.

2. Code Requirement - B13.20 - Interior Attachments Within the Beltline (B-N-2)

The ASME Section XI Code requires a VT-1 visual examination of accessible reactor interior surface attachment welds within the beltline each la-year interval. In the model 2 boiling water reactor, this includes the lower surveillauce specimen support bracket welds-to-vessel wall. In comparison, the BWRVIP requires the same examination method and frequency for the lower surveillance specimen support bracket welds.

3. Code Requirement - B13.30 - Interior Attachment Beyond the Beltline Region (B-N-2)

The ASME Section XI Code requires a VT-3 visual examination of accessible reactor interior surface attachment welds beyond the beltline each lO-year interval. In the model 2 boiling water reactor, this includes the upper surveillance specimen support bracket welds-to-vessel wall, the feedwater sparger support bracket welds-to-reactor vessel wall, the steam dryer support welds-to-reactor vessel wall, the guide rod support bracket weld-to-reactor vessel wall, and the shroud support plate-to-vessel weld. BWRVIP-48-A requires as a minimum the same VT-3 visual examination method as the Code for some of the interior attachment welds beyond the beltline region, and in some cases specifies an enhanced visual examination technique EVT-1 for these welds. For those interior attachment welds that have the same VT-3 method of visual examination, the same scope of examination (accessible welds), the same examination frequency (each lO-year interval) and ASME Section XI flaw evaluation criteria, the level of quality and safety provided by the BWRVIP requirements are equivalent to that provided by the ASME Code.

For the steam dryer support bracket attachment welds, the feedwater sparger support bracket attachment welds, and the shroud support plate-to-vessel welds, as applicable, the BWRVIP guidelines require an EVT-l visual examination at the same frequency as the Code. Therefore, the BWRVIP requirements provide the same level of quality and safety as that provided by the ASME Code.

The feedwater sparger bracket-to-vessel attachment weld is used as an example for comparison between the Code and BWRVIP examination requirements as discussed below.

Comparison to BWRVIP Requirements - Feedwater Sparger Bracket Welds (BWRVIP-48-A)

• The Code examination requirement is a VT-3 visual examination of each weld every 10 years.

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• The BWRVIP visual examination requirement is an EVT-l for the feedwater sparger bracket attachment welds with each weld examined every 10 years.

The BWRVIP visual examination method EVT-l has superior flaw detection and sizing capability, the examination frequency is the same as the Code requirements, and the same flaw evaluation criteria are used.

The Code VT-3 visual examination is conducted to detect component structural integrity by ensuring the components general condition is acceptable. An enhanced EVT-1 visual examination is conducted to detect discontinuities and imperfections on the examination surfaces, including such conditions as tight cracks caused by IGSCC or fatigue, and the relevant degradation mechanisms for BWR internal attachments.

Therefore, with the EVT-1 visual examination method, the same examination scope (accessible welds), the same examination frequency, and the same flaw evaluation criteria (ASME Section XI), the level of quality and safety required by the BWRVIP criteria is superior than that required by the Code.

4. Code Reqnirement - B13.40 - Welded Core Support Structures (B-N-2)

The ASME Code requires a VT-3 visual examination of accessible surfaces of the welded core support structure each lO-year interval. In the boiling water reactor, the welded core support structure has primarily been considered the shroud support structure, including the shroud. Historically, this requirement has been interpreted and satisfied differently across the industry. The proposed alternate examination replaces this ASME requirement with specific BWRVIP guidelines that examine susceptible locations for known relevant degradation mechanisms.

Comparison to BWRVIP Requirements - BWR Shroud Support (BWRVIP-38)

• The Code requires a VT-3 visual examination of accessible surfaces each la-year interval.

• The BWRVIP requires an enhanced visual examination technique (EVT-l) or volumetric examination (VT) every 10 years as compared to the Code requirement (VT-3).

BWRVIP recommended examinations of welded core support structures are focused on the known susceptible areas of this structure, including the welds and associated weld heat affected zones. In many locations, the BWRVIP guidelines require a volumetric examination of the susceptible welds at a frequency identical to the Code requirement.

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Comparison to BWRVIP Requirements - BWR Shroud (BWRVIP-76, Rev. I)

Shroud repair Tie-Rods have been installed at Oyster Creek Nuclear Generating Station.

Therefore BWRVIP-76, Rev. 1 requires inspection of the vertical shroud welds and the Tie-Rod repair hardware.

• The Code requires a VT-3 visual examination of accessible surfaces each IO-year intervaL

• The BWRVIP requires an enhanced visual examination technique (EVT-l) or volumetric examination (UT) of shroud vertical welds every lO-years minimum, as compared to the Code requirement (VT-3).

• The BWRVIP requires VT-3 and other appropriate techniques to examine the Tie-Rod repair hardware every ten years.

Therefore, the BWRVIP referenced examinations are the same or superior to the Code requirements. Shroud vertical welds and repair Tie-Rod examinations are recommended in BWRVIP-76-Rl and have the same basic VT-3 method of visual examination or better, the same examination frequency (each ten-year interval) and the same flaw evaluation criteria. Therefore, the BWRVIP requirements provide a level of quality and safety equivalent to that provided by the ASME Code.

For other welded core support structure components, the BWRVIP requires an EVT-l visual examination or UT of core support structures.

The BWRVIP recommended examinations specify locations that are known to be vulnerable to BWR relevant degradation mechanisms rather than accessible surfaces.

The BWRVIP examination methods (EVT-lor UT) are superior to the Code required VT-3 visual examination for flaw detection and characterization. The superior flaw detection and characterization capability, with an equivalent or more frequent examination frequency and the comparable flaw evaluation criteria, results in the BWRVIP criteria providing a level of quality and safety equivalent to or superior to that required by the Code requirements.

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Appendix 2 Reactor Vessel Inspection History Reactor Vessel Inspection History - Oyster Creek Nuclear Generatin2 Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Steam Dryer Fall 2010 Visual Re-inspected Steam Dryer indications and repair areas identified during previous outages and evaluated due to no change to use-as-is for one cycle.

Fall 2008 Visual Completed BWRVIP-139 required inspections of the ID of the dryer using GEH FireFly ROV. Identified 2 areas of fatigue cracking in drain channels and I area of fatigue cracking in support beam to mid-support ring weld. Evaluated as use-as-is for one cycle in accordance with BWRVIP-139 generic flaw evaluation.

Re-inspected Steam Dryer indications and repair areas identified during previous outages. Tie bar N-l lower repair area was found degraded and GEH issued JCO for one cycle of operation.

Fall 2006 Visual Re-inspect Steam Dryer Indications identified during previous outages.

EVT-l cracks in hold-down area from lR19.

VT-1 all 4 lifting lugs and EVT-1 indications on 135 deg. lug.

BWRVIP-139 required inspections (top side) completed. New fatigue indications were identified that required repair. Dryer repair project completed with 2 areas stop drilled and one crack in center baffle plate was cutout.

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Reactor Vessel Inspection History - Oyster Creek Nuclear Generating Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Core Shroud Fa1l2010 VT-3 VT-3 of2 Tie-Rods at 10 and 70 deg.

Locations (#2 and #8). No findings.

Fall 2008 UT / EVT-1 UT / EVT-1 inspection completed for all 10 shroud vertical welds. One indication found with UT in V10 weld - 1.76 inches long with depth of 0.47 inches. A technical evaluation was completed to use-as-is.

VT-3 Tie-Rods at 100 deg, 130 deg, 160 deg, 280 deg, and 350 deg. No findings.

Fall 2006 EVT-1 V-9 inspection of ID and OD. Two horizontal indications (transverse to the weld) were found adjacent to vertical weld on the ID surface. The indications were 2.75 and 1 inch in length and 30 and 35 inches above horizontal weld H5. A technical evaluation was completed to use-as-is.

VT-3 Tie-Rods at 170 deg, 220 deg and 310 deg. No findings.

VT-1 of Upper Bracket to Shroud Ledge interface on all 10 Tie-Rods. No findings.

Fall 2004 None No Examinations Required.

Fall 2002 None No Examinations Required.

Fall 2000 EVT-1 V-3, V-4, V-15 and V-16. This was a one sided exam from the OD. No findings.

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Reactor Vessel Inspection History - Oyster Creek Nuclear Generatin~ Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Core Shroud Fall 1998 UT V-7, V-8, V-lO and V-12.

(Continued) EVT-l V-ll1.D. Seven Tie-Rod assemblies baseline inspected.

V-I0 exhibited minor OD cracking away from the heat-affected zone. This cracking is believed to be associated with handling lugs removed after installation. All other inspected vertical welds were found free of indications.

With the inspections performed in 16R and 17R, all accessible vertical welds in the shroud core region are complete.

The following vertical welds could not be located. V-3, V-4, V-IS and V-I6.

Fall 1996 Visual Inspected per BWRVIP-07. Three often Tie-Rods inspected, no change from installation.

EVT-I, OD of V-9 and V-ll, (120" total). V-9 exhibited 3 small axial cracks in HAZ on the OD totaling 1.75". The ID of V-9 was free of axial cracks. A number of small transverse cracks were found on the OD and ID of V-9. V-II was free of any indications.

Analysis showed structural margin maintained.

Fall 1994 Ultrasonic Inspected per BWRVIP-OI and 03. Cracks and Visual were detected in the Shroud welds H2, H4, H6A, and H6B. Lack of fusion was detected in H3 weld and visual cracks on the ID surface.

The Tie-Rod modification was installed. Base line visual performed of the Tie-Rods.

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Reactor Vessel Inspection History - Oyster Creek Nuclear Generating Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Shroud Fall 2010 Visual EVT-1 of 6 of the 36 Lug / Clevis pin Support assemblies - #1, #2, #7, #8, #20 and #21.

No findings.

Fall 2008 Visual EVT-1 of 8 Lug / Clevis pin assemblies - #10,

1. 11, #14, #15, #17, #29, #30, and #36. No findings.

Fall 2006 Visual EVT-1 of 7 Lug / Clevis pin assemblies - #1,

1. 18, #19, #23, #24, #32 and #33.

Fall 2004 None No examinations required.

Fall 2002 UT 30% UT of H-9 from the OD (Drywell).

UT inspected H-9 weld in Nozzle N1A, N1C and N1E bioshield openings. Found one 4" long indication in the N1E nozzle area. This "service induced" indication is in the bottom side of the H9 weld and does not penetrate into the base metal of the RPV.

Fall 2000 Visual 25% ofH-9, cleaning performed and EVT-1 inspection completed. This completes 100%

inspection of the H-9 weld. No findings.

Fall 1998 Visual 25% of H-9, cleaning performed and enhanced VT-1, no findings Fall 1996 Visual 25% of H-9, (different area then the 1994 inspection), cleaning performed and enhanced VT-1, no findings.

Fall 1994 Visual 25% of H-9 cleaning performed and enhanced VT-1, no findings.

to CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page 17 of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generating Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Core Spray Fall 2010 Visual EVT-I of annulus piping fillet welds (all 10).

Piping No findings.

EVT-I of 25% shroud attachment welds -

Pipe Bracket PB 285 deg. No findings.

EVT-I of25% sample butt welds: P4aA, P4g1aA, P3aC, P3bC, P4dC, P4eC, P4fC, P4aB, P3aD, and P3bD. No findings.

Fall 2008 Visual EVT-I of annulus piping fillet welds (all 10).

No findings.

EVT-I of 25% shroud attachment welds -

Pipe Bracket PB 195 deg. No findings.

EVT-I of 25% sample butt welds: P3aA, P3bA,P2B,P4dD,P4eD,P4fl),P4gD,P4hD, P3aB, P3bB, and P4dB. No findings.

Fall 2006 Visual EVT-I of annulus piping fillet welds (all 10).

No findings.

EVT-I of 25% shroud attachment welds Pipe Bracket PB 103.5 deg. No findings.

EVT-I of 25% sample butt welds: P4bA, P4cA, P2A, P4g1aA, P4g/bA, P4hA, P4iC, P4g1aC, P4g/bC, P4hC,P4bB,P4eB,P4fB, P4gB and P4hB. No findings.

10 CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page 18 of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generatin~ Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Core Spray Fall 2004 Visual Accessible portions of the annulus piping Piping welds were cleaned using a nylon bmsh and (Continued) visual inspections performed utilizing the EVT-1 technique. All accessible portions of the following piping welds were visually inspected:

• L-3, L-3A, L-4, L-20A, L-13A, L-5, L-7, L-8, L-lO, L-ll, and L-12

• U-3, U-3A, U-4, U-15A, U-24A, U-7, U-8, U-9, U-lO, U-ll, U-12, U-16, and U-17 100% of annulus pipe brackets at 15°, 105°,

195° and 285°. No findings.

EVT-1 of all creviced welds in the annulus piping =U3, U3A, U4, U15A + U24A; L3, L3A, L4, L13A + L20A. No findings.

Fall 2002 Visual EVT-1 of a 25% sample (11 welds) of the butt welds (non-creviced) not inspected in 17R or 18R:

• Ul,UI5,UI7,UI8,UI9,U20

• Ll,L9,L13,LI6,L20 Inspect 100% of annulus pipe brackets (15°, 105°, 195° and 285°). No findings.

10 CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page 19 of32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generating Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Core Spray Visual Accessible portions of the annulus piping Fall 2000 Piping welds were cleaned using a nylon brush and (Continued) visual inspections performed utilizing the EVT-1 technique. All accessible portions of the following piping welds were visually inspected

• L3, L3A, L4, L6, LI3A, L14, L15 and L20A

• U3, U3A, U4, U7, U8 and U15A 100% of annulus pipe brackets 15°, 105° 195° and 285°. No findings.

Fall 1998 Visual All creviced welds in the annulus piping; sample (25%) of the non-creviced welds in the annulus piping:

• L2, L9, LI0, Lll, L12, L13, Ll7, L18, L19 and L20

• U2, U5, U6, U13, U14, U15, U21, U22, U23 and U24 Sample (25%) of pipe brackets 285°, 195°.

Fall 1996 Visual Inspected per BWRVIP-03. Cleaning of all accessible weld/HAZ surfaces and performed enhanced VT-1. No findings.

Fall 1994 Visual and Inspected VT-l, (l mil wire). No change to air test pinhole weld defect detected in slip joint in 1992. Note: Pinhole weld defect detected in 1992 in System 1. Analysis showed structural margin maintained.

10 CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page 20 of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generatin2 Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Core Spray Fall 2010 Visual EVT-l Sparger Pipe End Cap welds S4C 62 Sparger deg., S4C - 238 deg., S4D - 62 deg., and S4D

- 238 deg. No findings.

EVT-1 "T" box cover plate welds - SIC, S2C (LH), S2C (RH), SID, S2D (LH) and S2D (RH). No findings.

VT-1 spray nozzles - S3a, S3b, S3c C and D (CASS). No findings.

VT-1 of 50% of the sparger bracket welds -

SB-026 deg., SB-091 deg., SB-120 deg.,

SB-179 deg., SB-245 deg., SB-300 deg., and SB-359 deg. No findings.

Fall 2008 Visual EVT-1 Sparger Pipe End Cap welds:

S4A - 60 deg., S4A 240 deg., S4B - 60 deg.,

S4B - 240 deg. No findings.

EVT-l "T" box welds - SIA, S2A (LH), S2A (RH), SIB, S2B (LH) and S2B (RH). No findings.

VT-1 spray nozzles S3a, S3b, S3c-C. No findings.

VT-1 of 50% of the sparger bracket welds -

SB - 055, 065, 150, 208, 235, 271 and 330 deg. No findings.

10 CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page 21 of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generatin2 Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Core Spray Fall 2006 Visual EVT-l Sparger Pipe End Cap welds S4C - 60 Sparger deg., S4C - 240 deg., S40 - 60 deg., and S40 (Continued) - 240 deg. No findings.

EVT-1 "T" box welds - SIC, S2C (LH), S2C (RH), S10, S20 (LH) and S20 (RH). No findings.

VT-1 spray nozzles - S3a, S3b, S3c-B. No findings.

VT-1 of 50% of the sparger bracket welds -

SB - 026, 091, 120, 179,240,300, and 359 deg. No findings.

Fall 2004 Visual Inspected all sparger repair clamps. No findings.

Inspected end cap welds S4A-60, S4A-240, S4B-60, and S4B-240. No findings.

Inspected sparger brackets SB-055, 065, 150, 208,235,271 and 330. No findings Fall 2002 Visual and VT-1 all spargers, nozzles, end cap welds and Air Test repair clamps. No findings.

No new leaks were identified during the Air Test.

Fall 2000 Visual and All sparger end cap welds were cleaned and Air Test EVT-1 inspected. No findings.

VT-1 of spargers, repair clamps, and nozzles.

No findings.

No new leaks were identified during the Air Test.

10 CFR 50.55a RELIEF REQUEST 15R-Ol Revision 0 (Page of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generating Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Core Spray Fall 1998 Visual and All sparger repair clamps, both spargers.

Sparger Air Test (Continued)

Fall 1996 Visual and Inspected per BWRVIP-03. Cleaned end cap Air Test welds and performed enhanced VT -1. No findings. Tee box welds are clamped and not accessible to clean or visual. Performed VT-1, (1 mil wire), of sparger piping and nozzles. No findings.

Fall 1994 Visual and Performed VT-1, (1 mil wire) of sparger Air Test piping and nozzles. No findings.

1978-1980 Visual (2) Cracks in sparger piping. Repair clamps installed.

10 CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page 23 of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generating Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Top Guide Fall 2010 Visual EVT-ilnspection and Measurement of 3 Cracks in Grid Beams (#4, VT-3 & VT-6).

Some crack growth was observed. Flaw evaluation completed to support use as is.

BWRVlP-I83: Inspect 5% =7 cells. No findings.

Fall 2008 None Not required for this outage by analysis.

Fall 2006 Visual EVT-1 of selected known flaws in grid beams: #4, VT-3 and VT-6. One area showed no growth, while the other two had grown between 0.25" and 0.75" from the 2002 outage to the 2006 outage. A flaw evaluation was performed to use-as-is.

Fall 2004 Visual VT-1 of top guide hold down bolts at 303 and 123 degrees. No findings.

EVT-1 of VT-6 crack showed no measurable growth. Could not visually locate two other existing UT indications.

Fall 2002 Visual EVT-1 of two existing cracks measured in 18R outage (#3 and #5). No change to crack length identified.

10 CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page 24 of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generatin2 Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date ~Iethod Inspection Results, Repairs, Replacements, Scope Used Reinspections Top Guide Fall 2000 Visual Top guide hold down bolt assembly (Continued) VT-3 at 33° and 213°.

Top guide beam to rim fillet welds VT-1 at 33° and 213°. No findings.

VT-1 of two existing cracks (#3 and #5) with cleaning. Both cracks measured on both sides. Crack #5 showed approx. 1" growth.

Crack #3 showed no measurable growth.

Fa111998 None Not required for this outage by analysis.

Fall 1996 Ultrasonic 12 indications emanating from notches 100% grid detected at intersections of cross members. 5 beams of the 6 cracks on bottom side of member at mid span detected. Removed sample from beam with crack to investigate root case.

Fa111994 Visual [Under side of Top Guide] Three additional vertical cracks were detected at mid span locations. Disposition use as is.

Fa111992 Visual [Under side of Top Guide] Two additional vertical cracks were detected at mid span location. Disposition use as is.

Fa111991 Visual [Under side of Top Guide] A vertical crack was detected at mid span location.

Disposition use as is.

10 CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generating Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Core Plate Fall 2010 Visual VT-l inspected 3 In-core guide tube plugs 28-45,44-13 and 44-21. No findings.

Fall 2008 Visual Visually inspected core plate wedges at 96 0 and 276 0 azimuths. No findings.

Fa112006 None No exams were required.

Fall 2004 Visual No wedge inspections required. Inspected in-core guide tube plugs 04-29, 20-37, and 12-21. No findings.

Fa112002 Visual No inspections needed. Wedges replace hold down bolt inspections.

Fa112000 Visual Visually inspected all 8 wedges to verify integrity after first cycle of operation. All wedges found as installed.

Fall 1998 Visual Wedges installed. No further exams of core plate were performed.

Fall 1996 Visual Inspected top portion only of 18 hold down bolt that were not inspected in fall 1994 and top periphery section at bolt locations. No findings.

Fall 1994 Visual Inspected 18 hold down bolt tops only and top periphery at bolt locations inspected. No findings.

10 CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page 26 of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generating Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections SLC Fall 2010 VT-2 Inspected insulated nozzle from drywell. No pressure test leakage observed.

Fall 2008 VT-2 Inspected insulated nozzle from drywell. No pressure test leakage observed.

Fall 2006 UT PDI UT the Liquid Poison Nozzle N12! SE.

No findings.

Fall 2004 VT-2 Inspected insulated nozzle from drywell.

pressure test No leakage observed.

Fall 2002 Visual! PT PT of Liquid Poison Nozzle - no indications.

Inspect insulated nozzle from drywell during RPV pressure test. No leakage observed.

Fall 2000 VT-2 Inspected insulated nozzle from drywell.

pressure test No leakage observed.

Fall 1998 VT-2 during Not made accessible for direct exam.

Code pressure test Fall 1996 No Not made accessible.

Fall 1994 Inspection Perfonned

10 CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page 27 of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generating Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections CRD Guide Fall 2010 EVT-1, Inspected 2 guide tubes. Inspected 2 guide Tube VT-3 tube bases for CASS (VT-l). No findings.

Fall 2008 None Not required and not made accessible.

Fall 2006 EVT-1, Inspected 4 guide tubes. No findings.

VT-3 Fall 2004 EVT-l, Inspected 4 guide tubes. No findings.

VT-3 Fall 2002 EVT-l, Inspect 1 guide tube (46-43) removed to VT-3 support stub tube inspection. No findings.

Fall 2000 VT-l, VT-3 2 guide tubes. No findings.

Fall 1998 VT-3 15 guide tubes. No findings.

Fall 1996 No Not made accessible.

Inspection Performed Fall 1994 No Not made accessible.

Inspection Performed

10 GFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page 28 of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generating Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections CRD Stub Fall 2010 None No inspections required.

Tube Fall 2008 None No inspections required.

Fall 2006 None No inspections required.

Fall 2004 None No inspections required.

Fall 2002 VT-1 Visual inspection of 2 stub tubes found leaking at bottom head in Fall 2000 (42-43 and 46-39). No indications noted.

Fall 2000 VT-1 None made accessible.

VT-2 2 stub tubes found leaking at bottom head Pressure test (42-43 and 46-39). Performed UT of CRD housing to stub tube welds (J weld) and area of housing to be rolled. No indications. Roll repaired both leaking housings.

Fall 1998 No Not made accessible.

Fall 1996 Inspection Fall 1994 Performed

10 CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 of Reactor Vessel Inspection History - Oyster Creek Nndear Generating Station Components in Inspection Inspection Snmmarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections In-Core Fall 2010 No Not madeacclessl"'b1' e.

Housing Inspection Performed Fall 2008 No Not made accessible.

Inspection Performed Fall 2006 No Not made accessible.

Inspection Performed Fall 2004 No Not made accessible.

Inspection Performed Fall 2002 No Not made accessible.

Fall 2000 Inspection Fall 1998 Performed Fall 1996 Fall 1994

10 CFR 50.55a RELIEF REQUEST 15R-Ol Revision 0 (Page 30 of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generatin~ Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Scope Used Replacements, Reinspections Dry Tube Fall 2010 Replacement Replaced final 4 Dry tubes due to service life: IRM-15 and 16; SRM 22 and 23.

Fall 2008 Replacement Replaced 4 Dry tubes due to service life:

IRM-12, 13, 14 and SRM-21.

Fall 2006 Replacement Replaced 4 Dry tubes due to service life:

IRM-ll, 17, 18 and SRM-24.

Fall 2004 Visual VT-l of SRM 24 found tube not fully engaged in top guide. VT-1 of IRM 17 and IRM 18 found both tubes bowed.

Fall 2002 Visual No inspections required.

Fall 2000 Visual VT-1 five dry tubes. One found slightly bent

- use as is. No findings on others.

Fall 1998 Visual VT-lone dry tube, no findings Fall 1996 Visual VT-l one dry tube, no findings.

Fall 1994 Visual VT-1 four dry tubes, no findings.

Instrument Fall 2010 Visual VT-2 exam from vessel exterior.

Penetrations Fall 2008 No findings.

Fall 2006 Fall 2004 Fall 2002 Fall 2000 Fall 1998 Fall 1996 Fall 1994

10 CFR 50.55a RELIEF REQUEST I5R-Ol Revision 0 (Page 31 of 32)

Reactor Vessel Inspection History - Oyster Creek Nuclear Generating Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date l\lethod Inspection Results, Repairs, Replacements, Scope Used Reinspections VesselID Fall 2010 EVT-l EVT-l all feedwater sparger attachment wall Brackets bracket welds. No findings.

Fall 2008 None No inspection required.

Fall 2006 EVT-l EVT-1 both Guide Rod Brackets.

EVT-l all 3 Surv. Spec. Brackets.

No findings.

Fall 2004 EVT-l Inspected all 4 dryer support brackets.

No findings.

Fall 2002 EVT-l All feedwater sparger attachment brackets.

Both guide rod attachment brackets.

All surveillance sample brackets (30, 210 and 300 degree locations)

No indications on attachment welds.

Fall 2000 EVT-l All 4 dryer support brackets. Observed wear indications on brackets. No indications on attachment welds.

All feedwater attachment brackets inspected.

No indications on attachment welds.

Cracks observed on feedwater sparger to end bracket welds (non-safety-related component) on 2 ends.

Fall 1998 VT-l VT-1 of accessible portions of weld on guide Fall 1996 rod brackets, steam dryer brackets, Fall 1994 surveillance sample brackets. All attachment welds; no findings.

10 CFR 50.55a RELIEF REQUEST I5R~01 Revision 0 of Reactor Vessel Inspection History ~ Oyster Creek Nuclear Generatina Station Components in Inspection Inspection Summarize the Following Information:

BWRVIP Date Method Inspection Results, Repairs, Replacements, Scope Used Reinspections Fuel Snpport Fall 2010 Visual VT-1 inspection of 2 CRGT Bases (support Casting l) for CASS program.

No findings.

Fall 2008 None No inspection reQTIUired.

Fall 2006 Visual None inspected.

Fall 2004 Visual None inspected.

Fall 2002 Visual None inspected.

Fall 2000 Visual VT-3 (2) support casting. No findings.

Fall 1998 Visual VT-3 (24) support castings. No findings.

Fall 1996 Visual VT-3 (25) support castings. No findings.

Fall 1994 Visual VT-3 (17) support castings. No findings.

ReactorDM Fall 2010 UT Auto UT examined four (4) Category C nozzle to Welds safe end dissimilar metal (DM) welds (BWRVIP~75~ containing alloy 82/182. No findings.

A)

Fall 2008 UT Auto UT examined five (5) Category C nozzle to safe end dissimilar metal (DM) welds. One indication identified in NIA recirculation suction nozzle to safe end weld dispositioned as acceptable for 2 cycles in accordance with IWB-3600 flaw evaluation. The 0.21 inch ill connected indication was in the RPV nozzle to clad interface on the Reactor side of the Alloy 182 DM weld. The flaw evaluation was submitted to the NRC.

Note: All indications left "as-is" were analyzed and structural margins were acceptable for continued service.

Attachment 2 Relief Request I5R-02

10 CFR 50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 1 of 4)

Request for Relief from Pressure Testing the RPV Head Flange Seal Leak Detection Line In Accordance with 10 CFR 50.55a(a)(3)(ii)

1. ASME Code Component(s) Affected:

Code Class: 1

Reference:

Table IWB-2500-1 IWB-5200 Examination Category: B-P Item Number: B15.20

Description:

Pressure Testing the RPV Head Flange Seal Leak Detection Line Component Number: Class 1 RPV Head Flange Seal Leak Detection Line Drawing Number: BR 2002 Sh.l

2. Applicable Code Edition and Addenda:

The Oyster Creek Nuclear Generating Station lnservice Inspection (ISI) program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirement:

Table IWB-2500-1, Examination Category B-P, Item Number BI5.20, requires certain Class 1 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with Paragraph IWB-5220. This pressure test is to be conducted once per interval. The pressure retaining boundary for the test conducted at or near the end of each inspection interval shall be extended to all Class 1 pressure retaining components per Paragraph IWB-5222(b).

4. Reason for Request:

Exelon Generation Company, LLC (Exelon) is requesting a proposed alternative in accordance with 10 CFR 50.55a(a)(3)(ii) on the basis that compliance with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality.

The reactor vessel head flange leak detection line is separated from the reactor pressure boundary by one passive membrane, a silver-plated a-ring located on the vessel flange.

10 CFR 50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 2 of 4)

A second O-ring is located on the opposite side of the tap in the vessel flange. This line is required during plant operation and will indicate failure of the inner flange seal O-ring.

The configuration of this line precludes manual testing while the vessel head is removed.

The configuration of the vessel tap, combined with the small size of the tap and the high test pressure requirement prevents the tap from being temporarily plugged. Also, when the vessel head is installed, an adequate pressure test cannot be performed due to the fact that the inner O-ring is designed to withstand pressure in one direction only. Due to the groove that the O-ring sits in and the pin/wire clip assembly (See Figure I5R-02. I),

pressurization in the opposite direction into the recessed cavity and retainer clips would likely damage the O-ring and thus result in further damage to the O-ring.

Pressure testing of this line during the Class 1 System Leakage Test is precluded because the line will only be pressurized in the event of a failure of the inner O-ring. Purposely failing the inner O-ring to perform the Code required test would require purchasing a new set of O-rings, additional time and radiation exposure to de-tension the reactor vessel head, install the new O-rings, and then reset and re-tension the reactor vessel head. This is considered to impose a hardship and burden on the Oyster Creek Nuclear Generating Station.

Based on the above, the Oyster Creek Nuclear Generating Station requests relief from the ASME Section XI requirements for system leakage testing of the reactor vessel head flange seal leak detection line.

5. Proposed Alternative and Basis for Use:

A VT-2 visual examination on the Class 1 portion of the reactor pressure vessel flange leak detection line will be performed during each refueling outage when the RPV head is off and the head cavity is flooded above the vessel flange. The static head developed with the leak detection line filled with water will allow for the detection of any gross indications in the line. This examination will be performed each refueling outage as per the frequency specified by Table IWB-2500-1.

6. Duration of Proposed Alternative:

Relief is requested for the fifth ten-year lSI interval for Oyster Creek Nuclear Generating Station. The fifth ten-year lSI interval will begin on January 15,2013.

10 CFR 50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 3 of 4)

7. Precedents:

Similar relief requests have been approved for:

Limerick Generating Station, Units 1 and 2 Relief Request 13R-08 was approved in an NRC SER dated March 11, 2008.

Peach Bottom Atomic Power Station, Units 2 and 3 Relief Request I4R-25 was approved in an NRC SER dated February 26,2009.

Three Mile Island Nuclear Station, Unit 1 Relief Request I4R-03 was approved in an NRC SER dated July 20,2011.

10 CFR SO.SSa RELIEF REQUEST ISR-02 Revision 0 (Page 4 of 4)

FIGURE ISR-02.1 O-RING CONFIGURATION FLANGE LEAK-OFF LINE SECTION A

Attachment 3 Relief Request 15R-05

10 CFR 50.55a RELIEF REQUEST I5R-05 Revision 0 (Page I of 2)

Request for Relief for Expanded Applicability for use of ASME Code Case N-661-1, Alternative Requirements for Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service In Accordance with 10 CFR 50.55a(a)(3)(O

1. ASME Code Component(s) Affected:

Code Class: 2 and 3

Reference:

IWA-2441(b) and ASME Code Case N-661-1 Examination Category: NA Item Number: NA

Description:

Expanded Applicability for use of ASME Code Case N-661-1, "Alternative Requirements for Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service" Component Number: Class 2 and 3 Piping

2. Applicable Code Edition and Addenda

The Oyster Creek Nuclear Generating Station lnservice Inspection (lSI) program is based on the American Society of Mechanical Engineers (AS ME) Boiler and Pressure Vessel (B&PV) Code.Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirement

IWA-244 I (b) requires Code Cases be applicable to the Edition and Addenda specified in the Inspection Plan.

ASME Code Case N-661-1, "Alternative Requirements for Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service," provides requirements that may be used to restore wall thickness for raw water piping systems that have experienced internal wall thinning.

4. Reason for Request

Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

10 CFR 50.55a RELIEF REQUEST I5R-05 Revision 0 (Page 2 of 2)

On January 15,2013 Oyster Creek Nuclear Generating Station will start its fifth ten-year interval lSI program under the requirements of the 2007 Edition through the 2008 Addenda of ASME Section XI. When implementing this edition of ASME Section XI, Paragraph IWA-2441(b) requires code cases be applicable to the Edition and Addenda specified in the Inspection Plan.

ASME Code Case N-661 1 has an applicability limited up to the 2004 Edition through the 2005 Addenda, which is identified in the latest applicability index for ASME Section XI Code Cases. Since ASME Code Case N-661-1 only applies up to the 2004 Edition through the 2005 Addenda, Paragraph IWA-2441 (b) does not allow the use of ASME Code Case N-661-1 for the Oyster Creek Nuclear Generating Station fifth ten-year interval lSI program.

5. Proposed Alternative and Basis for Use:

Oyster Creek Nuclear Generating Station requests the applicability of ASME Code Case N-661-1 be extended to the 2007 Edition through the 2008 Addenda for use in the plant's fifth interval lSI program. The NRC has accepted the use of ASME Code Case N-661-1 as an acceptable method for restoring wall thickness for raw water piping systems that have experienced internal wall thinning in the latest revision of Regulatory Guide 1.147, Revision 16.

No technical changes to ASME Code Case N-661-1 are being proposed in this relief request. This relief request is being submitted to correct a timing situation, which has resulted from the application of the 2007 Edition through the 2008 Addenda of ASME Section XI for Oyster Creek Nuclear Generating Station. Since no technical change is proposed in this relief request, Oyster Creek Nuclear Generating Station considers that this alternative provides an acceptable level of quality and safety, and is consistent with provisions of 10 CFR 50.55a(a)(3)(i).

6. Duration of Proposed Alternative:

Relief is requested for the fifth ten-year lSI interval for Oyster Creek Nuclear Generating Station. The fifth ten-year lSI interval will begin on January 15,2013.

7. Precedents:

None.

Attachment 4 Relief Request I5R-06

10 CFR 50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 10f2)

Request for Relief for Continuous Pressure IVlonitoring of the Control Rod Drive (CRn) System Accumulators In Accordance with 10 CFR 50.55a(a)(3)(i)

1. ASl\1E Code Component(s) Affected:

Code Class: 2

Reference:

Table IWC-2500-1 Examination Category: C-H Item Number: C7.10

Description:

Continuous Pressure Monitoring of the Control Rod Drive (CRD) System Accumulators Component Number: CRD Accumulators and Associated Piping Drawing Number: GE197E871 Sh. 1

2. Applicable Code Edition and Addenda

The Oyster Creek Nuclear Generating Station Inservice Inspection (lSI) program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirement

Table IWC-2500-1, Examination Category C-H, Item Number C7.1O, requires all Class 2 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with IWC-5220. This pressure test is to be conducted once each inspection period.

4. Reason for Request

Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative provides an acceptable level of quality and safety.

As required by an Oyster Creek Nuclear Generating Station operating procedure, the Control Rod Drive (CRD) system accumulator pressure must be greater than or equal to 940 psig to be considered operable. The accumulator pressure is continuously monitored by system instrumentation. Since the accumulators are isolated from the source of make-up nitrogen, the continuous monitoring of the CRD accumulators functions as a pressure decay type test. When an accumulator trouble alarm is received in the control room, action is taken to repair the accumulator in accordance with the Oyster Creek Nuclear

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Generating Station operating procedure. If the accumulator cannot be repaired within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, then the control rod is declared inoperable.

Since monitoring the nitrogen side of the accumulators is continuous, any leakage from the accumulator would be detected by normal system instrumentation. An additional VT-2 visual examination performed once per inspection period would not provide an increase in safety, system reliability, or structural integrity. In addition, performance of a VT-2 visual examination would require applying a leak detection solution to 137 accumulators resulting in additional radiation exposure without any added benefit in safety. This inspection would thus not be consistent with ALARA practices.

Relief is requested from the VT-2 visual examination requirements specified in Table IWC-2500-I for the nitrogen side of the CRD system accumulators on the basis that Oyster Creek Nuclear Generating Station operating procedure exceeds the code requirement for a VT-2 visual examination.

5. Proposed Alternate and Basis for Use:

As an alternate to the VT-2 visual examination requirements of Table IWC-2500-1, Oyster Creek Nuclear Generating Station will perform continuous pressure decay monitoring and corrective actions as discussed in the Oyster Creek Nuclear Generating Station operating procedure for the nitrogen side of the CRD accumulators including attached piping.

6. Duration of Proposed Alternative:

Relief is requested for the fifth ten-year lSI interval for Oyster Creek Nuclear Generating Station. The fifth ten-year lSI interval will begin on January 15,2013.

7. Precedents:

Similar relief requests have been approved for:

Dresden Nuclear Power Station, Units 2 and 3 Relief Request I4R-07 was approved in an NRC SER dated September 4,2003.

Quad Cities Nuclear Power Station, Units 1 and 2 Relief Request I4R-06 was approved in an NRC SER dated January 28,2004.

LaSalle County Station, Units 1 and 2 Relief Request I3R-09 was approved in an NRC SER dated January 30, 2008.

Attachment 5 Relief Request I5R-07

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Request for Relief for Expanded Applicability for use of ASME Code Case N-532-4, RepairlReplacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission In Accordance witb 10 CFR 50.55a(a)(3)(i)

1. ASME Code Component(s) Affected:

Code Class: All Classes

Reference:

IWA-2441(b) and ASME Code Case N-532-4 Examination Category: NA Item Number: NA

Description:

Expanded Applicability for use of ASME Code Case N-532-4, "RepairlReplacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission" Component Number: NA

2. Applicable Code Edition and Addenda

The Oyster Creek Nuclear Generating Station Inservice Inspection (lSI) program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirement

IWA-2441(b) requires Code Cases be applicable to the Edition and Addenda specified in the Inspection Plan.

ASME Code Case N-532-4, "RepairlReplacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission," provides requirements that may be used to document repair/replacement activities.

4. Reason for Request

Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

On January 15,2013 Oyster Creek Nuclear Generating Station will start its fifth ten-year interval lSI program under the requirements of the 2007 Edition through the 2008 Addenda of ASME Section XI. When implementing this edition of ASME Section XI,

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Paragraph IWA-244I(b) requires code cases be applicable to the Edition and Addenda specified in the Inspection Plan.

ASME Code Case N-532-4 has an applicability limited up to the 2004 Edition through the 2005 Addenda, which is identified in the latest applicability index for ASME Section XI Code Cases. Since ASME Code Case N-532-4 only applies up to the 2004 Edition through the 2005 Addenda, Paragraph IWA-2441(b) does not allow the use of ASME Code Case N-532-4 for the Oyster Creek Nuclear Generating Station fifth ten-year interval lSI program.

5. Proposed Alternative and Basis for Use:

Oyster Creek Nuclear Generating Station requests the applicability of ASME Code Case N-532-4 be extended to the 2007 Edition through the 2008 Addenda for use in the plant's fifth interval lSI program. The NRC has accepted the use of ASME Code Case N-532-4 as an acceptable method for repair/replacement activity documentation requirements and inservice summary report preparation and submission in the latest revision of Regulatory Guide 1.147, Revision 16.

No technical changes to ASME Code Case N-532-4 are being proposed in this relief request. This relief request is being submitted to correct a timing situation, which has resulted from the application of the 2007 Edition through the 2008 Addenda of ASME Section XI for Oyster Creek Nuclear Generating Station. Since no technical change is proposed in this relief request, Oyster Creek Nuclear Generating Station considers that this alternative provides an acceptable level of quality and safety, and is consistent with provisions of 10 CFR 50.55a(a)(3)(i).

6. Duration of Proposed Alternative:

Relief is requested for the fifth ten-year lSI interval for Oyster Creek Nuclear Generating Station. The fifth ten-year lSI interval will begin on January 15,2013.

7. Precedents:

None.