ML13169A062

Relief from the Requirements of the ASME Code, Relief Request No. I5R-01

ML13169A062
Person / Time
Site:	Oyster Creek
Issue date:	08/05/2013
From:	Veronica Rodriguez Plant Licensing Branch 1
To:	Pacilio M Exelon Nuclear
Lamb J
References
TAC ME9490
Download: ML13169A062 (13)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555"()001 August 5,2013 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60SS5

SUBJECT:

OYSTER CREEK NUCLEAR GENERATING STATION - RELIEF FROM THE REQUIREMENTS OF THE ASME CODE, RELIEF REQUEST NO. 15R-01 (TAC NO. ME9490)

Dear Mr. Pacilio:

By letter dated August 28,2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12243A287), as supplemented by letter dated March 8,2013 (ADAMS Accession No. ML13071A092), Exelon Generation Company (Exelon, the licensee),

submitted Request for Relief (RR) 15R-01 for the U.S. Nuclear Regulatory Commission's (NRC's) approval. The licensee proposed an alternative to certain requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),

Section XI. RR ISR-01 is requested for the fifth 10-year inservice inspection (lSI) interval at Oyster Creek Nuclear Generating Station (Oyster Creek), which commenced on January 1S, 2013. The 2007 Edition through the 2008 Addenda of the ASME Code,Section XI, is the current Code of record at Oyster Creek. Specifically, pursuant to Title 10 of the Code Federal Regulations (10 CFR), paragraph SO.SSa(a)(3)(i), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety.

The NRC staff has reviewed the subject request and has concluded, as set forth in the enclosed safety evaluation, that the proposed alternative described in RR ISR-01 provides an acceptable level of quality and safety. Accordingly, the staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i), and is in compliance with the ASME Code's requirements.

Therefore, the NRC staff authorizes the alternative described in RR ISR-01 for the fifth lSI interval at Oyster Creek, which began on January 15, 2013, and ends on January 14, 2023. All other ASME Code requirements for which relief was not specifically requested and approved in the subject request remain applicable, including a third party review by the Authorized Nuclear Inservice Inspector.

M. Pacilio -2 If you have any questions regarding this matter, please contact the Senior Project Manager, John G. Lamb at (301) 415-3100 or bye-mail at John.Lamb@nrc.gov.

Sincerely,

~

Plant ROdr;~

LiCenSingg~~:~~

Veronica . Chief Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-219

Enclosure:

Safety Evaluation cc w/enclosure: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF 15R-01 FOR THE FIFTH 10-YEAR INSERVICE INSPECTION INTERVAL OYSTER CREEK NUCLEAR GENERATING STATION EXELON NUCLEAR DOCKET NO. 50-219

1.0 INTRODUCTION

By letter dated August 28. 2012, (Agencywide Documents Access and Management System (ADAMS). Accession No. ML12243A287), as supplemented March 8,2013 (ADAMS Accession No. ML13071A092), Exelon Generation Company (Exelon, the licensee) proposed Requestfor Relief (RR) No. 15R-01 from certain requirements of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code), under the provisions of Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.55a(a)(3)(i), for the fifth 10-year Inservice Inspection (lSI) Program for Oyster Creek Nuclear Generating Station (OCNGS).

2.0 REGULATORY EVALUATION

lSI of the ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code, and applicable addenda, as required by paragraph 50.55a(g) of 10 CFR, except where specific relief has been granted by the Commission pursuant to 10 CFR SO.SSa(g)(6)(i). The regulations in 10 CFR SO.SSa(a)(3), state, in part, that alternatives to the requirements of paragraph (g) may be used, when authorized by the U.S. Nuclear Regulatory Commission (NRC), if the licensee demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The licensee has proposed alternatives from ASME Code requirements pursuant to 10 CFR 50.55a(a)(3)(i). The ASME Code of record for the OCNGS fifth 10-year lSI interval program is the 2007 Edition through the 2008 Addenda of Section XI of the ASME Code. The fifth 10-year lSI interval began on January 15, 2013, and is projected to end on January 14, 2023.

Enclosure

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3.0 TECHNICAL EVALUATION

RR 15R-01 3.1 ASME Code Component(s) Affected (as stated by the licensee)

Code Class: 1

Reference:

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

Description:

Use of BWRVIP [Boiling-Water Reactor Vessel Internal Project] 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 Beltlone Region. and Core Support Structure 3.2 ASME Code Requirements (as stated by the licensee)

ASME [Code],Section XI requires the examination of components within the reactor pressure vessel. These examinations are included in [ASME Code.

Section XI,] Table IWB-2500-1, Categories B-N-1 and B-N-2 and identified with the following [ASME Code,Section XI,] Item Numbers:

B13.10 Examine accessible areas of the reactor vessel interior each period by the [Visual Testing (VT)]-3 visual examination method (B-N-1 ).

B13.20 Examine interior attachment welds within the beltline region each interval by the VT-1 visual examination method (B-N-2).

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.

In its request for additional information (RAI), dated February 5,2013 (ADAMS Accession No. ML13014A570), the NRC staff inquired about the shroud support leg weld (H12) which requires inspection per the requirements specified in the NRC staff's Safety Evaluation (SE) for the BWRVIP-38 report, "BWR Vessel Internal Project. BWR Shroud Support Inspection and Flaw Evaluation Guidelines." In its response, dated March 8, 2013, the licensee stated that:

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[The OCNGS] is a GE [General Electric Co.] BWR-2 product line designed vessel. As such, the shroud support structure consists of a cone, with its base welded to the Reactor Vessel at the H-9 weld, and the top supporting the core shroud. There are no shroud support legs associated with this support structure.

Therefore, there are no H-12 welds in the OCNGS vessel.

In its RAI, the NRC staff requested that the licensee identify the welds that were made with Alloy 182 weld electrode in OCNGS Reactor Vessel Internals (RVI) components. Since Alloy 182 welds are prone to intergranular stress-corrosion cracking, the licensee was requested to provide the history of inspections and the results of these inspections of these welds, and subsequent examination criteria, if cracking was found in these welds. In its response, the licensee stated that:

With regards to Examination Category B-N-1, there are no Alloy 182 welds. With regards to Examination Category B-N-2, the following locations have Alloy 182 welds:

1. Steam Dryer Support Brackets

2. Guide Rod Attachment Brackets

3. Feedwater Sparger Attachment Brackets

4. Surveillance Sample Brackets

5. Shroud Support Lug I Clevis Assemblies

6. Conical Shroud Support-to-RPV [reactor pressure vessel] Attachment Weld H-9

7. Conical Shroud Support Radial Welds The results of the inspections for items 1 through 4 are contained in Appendix 2[1] of the Relief Request, page 31 of 32 ("Vessel 10 Brackets"). The inspections of the Shroud Support Lug I Clevis Assemblies (item 5) are contained in Appendix 2 of the Relief Request, page 16 of 32 ("Shroud Support").

The results of the inspections of the H-9 weld (item 6) are contained in Appendix 2 of the Relief Request, page 16 of 32 ("Shroud Support"). We note that in the 2012 refueling outage, the H-9 weld was re-inspected with improved UT

[ultrasonic testing] inspection techniques. The results of this inspection identified no significant change to the recordable indication found in the previous inspection results from the fall 2002 outage. For the Conical Shroud Support Radial Welds (item 7), neither the ASME Code nor the BWRVIP Guidelines contain specific inspection requirements for these welds.

The NRC staff reviewed the inspection results for items 1 through 4 in Appendix 2 and found that the licensee had no findings. For the inspections of the Shroud Support Lug/Clevis Assemblies (item 5) contained in Appendix 2, the licensee had no findings except for H-9 (item

6) weld, which is addressed above by the licensee. As noted above by the licensee, the Conical 1 Appendix 2 of the Relief Request, page 31 of 32 ("Vessel 10 Brackets") is not provided in this safety evaluation (SE) and is contained in the licensee's submittal dated March 8, 2013.

-4 Shroud Support Radial Welds (item 7), neither the ASME Code nor the BWRVIP Guidelines contain specific inspection requirements for these welds.

3.3 Licensee's Basis for Relief Request (as stated by the licensee)

Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested for the proposed alternative to the [ASME} 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 BWRV1P 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 [ASME Code}

inspection requirements were prepared before the BWRVIP 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.

3.4 Licensee's Proposed Alternative Examination (as stated by the licensee)

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

[OCNGS} will satisfy the Examination Category B-N-1 and B-N-2 requirements as described in Table 1 in accordance with BWRVIP 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:

BWRVIP-03, "BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines" BWRVIP-18, Revision 1, "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines" BWRVIP-25, "BWR Core Plate Inspection and Flaw Evaluation Guidelines" BWRVIP-26-A, "BWR Top Guide Inspection and Flaw Evaluation Guidelines"

- 5 BWRVIP-27 -A, "BWR Standby Liquid Control System/Core Plate l\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 10 Attachment Weld Inspection and Flaw Evaluation Guidelines" BWRVIP-76, Revision 1, "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 [Code,Section XI,] Examination Category B-N 1 and B-N-2 requirements with the above current BWRVIP guideline requirements, as applicable, to the [OCNGS BWR Type 2 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 principle, and is expected to issue Safety Evaluations for many of these guidelines. .. [Note: "in principle" means that, for some reports, final SEs have been written with the exception of BWRVIP-03 and BWRVIP-94 reports, but the final BWRVIP acceptance reports which incorporate these SEs for some of the reports may not have been issued]. Therefore, use of these guidelines, as an alternative to the subject [ASME] 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 12 ("Comparison of Code Examination Requirements to BWRVIP Examination Requirements") provides specific examples which compare the inspection requirements of [ASME Code,Section XI, Table IWB-2500-1, Categories B-N-1 and B-N-2, Item Numbers B13.10, B13.20, B13.30, and B 13.40,] to the inspection requirements in the 2 Appendix 1 is not contained in this SE and may be found in the licensee's letter dated March 8, 2013.

-6 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 [ASME] Code requirements, provides an acceptable level of quality and safety and will not adversely impact the health and safety of the public.

3.5 NRC Staff Evaluation The ASME Code requires 100 percent visual examination on all accessible areas or surfaces, as applicable, of reactor vessel interior and core support structures, respectively. The NRC staff has reviewed the licensee's submittal listing the BWRVIP inspection guidelines that have been proposed as alternatives to the ASME Code requirements given above. The NRC has previously reviewed, approved, and issued SEs on all BWRVIP guidelines listed in the licensee's request. It has also been verified that no NRC conditions have been imposed on the use of the BWRVIP guidelines in the approved SEs.

For ASME Code Item Number B 13.10, reactor vessel interior accessible areas, the applicable BWRVIP guidelines require a visual VT-3 examination of reactor vessel accessible areas, which are defined as the spaces above and below the core made accessible during normal refueling outages at a more frequent basis than that required by the ASME Section XI Code.

Examination of core spray piping and spargers (BWRVIP-18-R1), top guide (BWRVIP-26-A),

interior attachments (BWRVIP-48-A), core shroud welds (BWRVIP-76-R1), and shroud support (BWRVIP-38) provides such access. The licensee, with remote camera systems, will perform an equivalent VT -3 visual examination of the reactor vessel areas noted above, below when accessible in the surrounding core annulus area.

These examinations will provide an equivalent method of visual VT-3 examination on a more frequent basis than that required by the ASME Code,Section XI. The licensee will look for evidence of wear, structural degradation, loose, missing, or displaced parts, foreign materials, and corrosion product buildup during the course of implementing the subject BWRVIP examination requirements. Therefore, the specified BWRVIP Guideline requirements meet or exceed the subject ASME Code requirements for examination method and frequency of the interior of the reactor.

For ASME Code Item Number B13.20, interior attachments within the beltline, the ASME Code requires a VT-1 visual examination of accessible reactor interior surface attachment welds within the beltline each 10-year lSI interval. In a BWR Type 2 reactor the examinations include the lower surveillance specimen support bracket welds-to-vessel wall. In comparison, the BWRVIP-48-A, Table 3-2 requires the same examination method and frequency for the lower surveillance specimen support bracket welds. Therefore, the BWRVIP Guideline requirements meet the subject ASME Code for examination method and frequency of the interior attachments.

For ASME Code Item Number B13.30, interior attachments beyond the beltline region, the ASME Code requires a VT-3 visual examination of accessible reactor interior surface attachment welds beyond the beltline each 10-year lSI interval. In a BWR Type 2 reactor, the examinations include 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

-7 to-reactor vessel wall, the guide rod support bracket weld-to-reactor vessel wall, and the shroud support to vessel weld (H9). BWRVIP-48-A requires a VT-3 visual examination as the ASME Code for the interior attachment welds beyond the beltline region, and in some cases specifies an enhanced visual examination technique (EVT-1) for the subject component interior attachment welds. For the subject interior attachment welds, the level of quality and safety provided by the BWRVIP-48-A requirements are equivalent to the ASME Code,Section XI requirements.

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-48-A guidelines require an EVT-1 visual or UT(shroud support plate) examination at the same frequency as the ASME Code, Section XI-requirements. Therefore, the BWRVIP-48-A requirements provide an equivalent level of quality and safety to the ASME Code, Section XI-requirements.

For ASME Code Item B13.40, Welded Core Support Structures, the ASME Code requires a VT 3 visual examination of accessible surfaces of the welded core support structure each 10-year lSI interval. In the BWR, the welded core support structure has primarily been considered the shroud support structure, including the shroud. The BWRVIP-38 proposed alternate examination replaces this ASME Code-requirement with specific guidelines that examine susceptible locations for known relevant degradation mechanisms. The BWRVIP-38 guidelines requires an EVT-1 visual or UT examination every 10 years as compared to the ASME Code requirement of a VT-3 visual examination. Based on the above evaluation, the BWRVIP-38 guidelines meet or exceed the ASME Code requirements for the examination method and frequencies of the RPV interior and core support structures.

Based on the above basis provided by the licensee, the NRC staff has determined that the licensee's proposed alternatives to use the subject BWRVIP guidelines summarized in the licensee's submittal provide an acceptable level of quality and safety.

3.6 Duration of Relief RR 15R-01 is requested for the fifth 1O-year lSI interval of OCNGS, which commenced on January 15, 2013.

4.0 CONCLUSION

As set forth above, the NRC staff has determined that authorizing alternatives pursuant to 10 CFR 50.55a(a)(3)(i) for 15R-01 provides an acceptable level of quality and safety.

Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i). Therefore, the NRC staff authorizes the alternatives to the specified ASME Code,Section XI, contained in RR No. 15R-01 for the duration of the OCNGS fifth 10-year lSI interval.

-8 All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

Principle Contributor: T. McLellan Date: August 5, 2013

10 CFR 50.55a RELIEF REQUEST 15R-01 TABLE 1: Comparison of ASME Examination Category B-N-1 and B-N-2 Requirements With BWRVIP Guidance Requirements(1)

I ASME Code Component ASME Exam ASME ASME Authorized BWRVIP BWRVIP BWRVIP Item Scope Code Code Alternative Exam Exam Frequency Number, Exam Frequency Exam Scope Table IWB-2500-1 I B13.10 Reactor Vessel Interior Accessible VT-3 Each BWRVIP-18-R1, Overview examinations of components during Areas period 25, 26-A, 27-A, BWRVIP examinations satisfy Code VT-3 38, 47-A,48-A, visual inspection requirements.

and 76-R1 r-~ --- I B 13.20 Interior Attachment Accessible VT-1 Each BWRVIP-48-A, Bracket VT-1 Each 10-year Within Beltline Region Welds 10-year Table 3-2 Attachment Interval

- Lower Surveillance Interval Specimen Holder

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Brackets  !

B 13.30 Interior Attachments Accessible VT-3 Each Beyond Beltline Welds 10-year ~~~ ~ --- I Guide Rod Brackets Interval BWRVIP-48-A, Bracket VT-3 Each 10- year

~able3-2 Attachment Interval Steam Dryer BWRVIP-48-A, Bracket EVT-1 Each 10-year Support Brackets Table 3-2 Attachment Interval Feed water BWRVIP-48-A, Bracket EVT-1 Each 10-year Sparger Brackets Table 3-2 Attachment Interval Upper Surveillance BWRVIP-48-A, Bracket VT-3 Each 10-year Specimen Holder Table 3-2 Attachment interval Brackets Shroud Support BWRVIP-38, Wfid HgI<!J EVT-1 or UT Maximum of (Weld H9) 3.1.3.2, 6 years for one-Figure 3-5 sided EVT-1, Maximum of 10

~-~

years for UT Attachment

-2 10 CFR 5O.55a RELIEF REQUEST 15R-01 TABLE 1 Comparison of ASME Examination Category B-N-1 and B-N-2 Requirements With BWRVIP Guidance Requirements I

ASME Code I ASME ASME BWRVIP Item ASMECode Authorized BWRVIP BWRVIP Component Code Code Exam Number, Exam Scope Alternative Exam Frequency Exam Frequency Scope

~ ~-

Table B13.4O Welded Core Accessible VT-3 Each 1 BWRVIP-38, Shroud EVT-1 or UT Based on as found Support Structure Surfaces ~year 3.1.3.2, Support conditions, to a Shroud Support Interval Figure 3-5 maximum 6 years for one-sided EVT 1, 10 years for UT where accessible Shroud Vertical Welds BWRVIP-76 Vertical and EVT-1 or UT Maximum 6 years R1, Ring for one-sided EVT Section 3.3, Segment 1, 10 years for UT Figure 3-1 Welds as applicable BWRVIP-76 Shroud Repairs\'>}

R1, VT-3 Tie-Rod Per designer Section 3.5 Repair recommendation per BWRVIP-76-R1 I

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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 appropriate 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.

M. Pacilio -2 If you have any questions regarding this matter, please contact the Senior Project Manager, John G. Lamb at (301) 415-3100 or bye-mail at John.Lamb@nrc.gov.

Sincerely, Ira!

Veronica Rodriguez, Acting Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-219

Enclosure:

Safety Evaluation cc w/enclosure: Distribution via Listserv DISTRI BUTION:

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