ML101190527

Relief Request 13R-06 for Detailed Visual Examination During Appendix J Pnuematic Leakage Testing

ML101190527
Person / Time
Site:	Braidwood
Issue date:	05/11/2010
From:	Shawn Campbell Plant Licensing Branch III
To:	Pardee C Exelon Generation Co, Exelon Nuclear
David Marshall NRR/DORL/LPL3-2 415-154
References
TAC ME2194, TAC ME2195
Download: ML101190527 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 11,2010 Mr. Charles G. Pardee President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2 - RELIEF REQUEST 13R-06 FOR DETAILED VISUAL EXAMINATION DURING APPENDIX J PNUEMATIC LEAKAGE TESTING (TAC NOS. ME2194 AND ME2195)

Dear Mr. Pardee:

By letter to the Nuclear Regulatory Commission (NRC), dated September 4, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092510267), Exelon Generation Company, LLC (the licensee) submitted Relief Request (RR) 13R-06 for relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Subsection IWE, for the third 10-year inservice inspection (lSI) interval for Braidwood Station (Braidwood), Units 1 and 2. The third 10-year lSI interval for Braidwood is currently scheduled to end on July 28,2018, for Unit 1 and on October 16, 2018, for Unit 2. In accordance with Title 10 of the Code of Federal Regulations (10 CFR), 50.55a(a)(3)(ii),

the licensee requested relief from the ASME Code,Section XI, requirements in Subarticle IWE 5240 to perform certain detailed visual examinations during 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," pneumatic leakage testing.

The NRC staff has reviewed the licensee's submittal and has determined that compliance with the specified code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, and that the alternative proposed in RR 13R-06 will provide reasonable assurance of structural integrity and leak tightness of the affected components. Therefore, the alternative proposed in RR 13-06 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii), for the third 1O-year lSI interval for Braidwood, Units 1 and 2. All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third party review by the Authorized Nuclear Inservice Inspector. The NRC staffs safety evaluation is enclosed.

C. Pardee -2 Please contact Mr. Marshall David at (301) 415-1547 if you have any questions on this action.

Sincerely,

~:).~

Stephen J. Campbell, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. STN 50-456 and 50-457

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON,1l.C.-20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO. 13R-06 EXELON GENERATION COMPANY, LLC.

BRAIDWOOD STATION, UNITS 1 AND 2 DOCKET NOS. STN 50-456 AND 50-457

1.0 INTRODUCTION

By letter to the Nuclear Regulatory Commission (NRC), dated September 4, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092510267, Reference 1), Exelon Generation Company, LLC (EGC, the licensee) submitted Relief Request (RR) 13R-06 for relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Subsection IWE, for the third 10-year inservice inspection (lSI) interval for Braidwood Station (Braidwood), Units 1 and 2.

The third 1O-year lSI interval for Braidwood is currently scheduled to end on July 28, 2018, for Unit 1, and on October 16, 2018, for Unit 2. In accordance with Title 10 of the Code of Federal Regulations (10 CFR), 50.55a(a)(3)(ii), the licensee requested relief from the ASME Code,Section XI, requirements in Subarticle IWE-5240 to perform certain detailed visual examinations during 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors," pneumatic leakage testing.

The NRC staff has reviewed and evaluated the information provided by the licensee in its submittal. The results of the NRC staff's review are presented in the remainder of this safety evaluation.

2.0 REGULATORY EVALUATION

The regulations in 10 CFR 50.55a, "Code and standards," address the use of codes and standards as they relate to structures, systems, and components, which must be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed. Specifically, 10 CFR 50.55a(b)(2)(ix) identifies the regulatory conditions that apply to the use of Subsection IWE of the ASME Code for the examination of metal containments and the liners of concrete containments, while 10 CFR 50.55a(g) delineates the requirements for lSI of those components (including supports),

which are classified as ASME Code Class 1, Class 2, or Class 3. According to 10 CFR 50.55a(a)(3), proposed alternatives to the requirements of paragraphs (c), (d), (e), (f), (g) and (h) of 10 CFR 50.55a, or portions thereof, may be used when authorized by NRC's Director of the Office of Nuclear Reactor Regulation provided the applicant demonstrates that: (i) the proposed

-2 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. In RR 13R-06, the licensee requested relief from certain ASME Code,Section XI, Subsection IWE requirements pursuant to 10 CFR 50.55a(a)(3)(ii).

On September 22, 2008, the NRC staff authorized a similar relief request relating to the detailed visual examination requirements of Subsection IWE of the ASME Code for the second containment lSI Interval for the Duane Arnold Energy Center (ADAMS Accession No.

I\IIL082460235, Reference 2).

3.0 TECHNICAL EVALUATION

3.1 Description of RR 13R-06 In Section 4.0 of RR 13R-06 in Reference 1, the licensee requested relief, pursuant to 10 CFR 50.55a(a)(3)(ii), from the requirements of Subarticle IWE-5240 to perform a detailed visual examination during the Appendix J local leak rate test for replacements (installed by mechanical connection) or minor repair activities. This request was submitted on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

In the RR, the licensee stated that, per IWE-2310(d), detailed visual examinations must be performed in accordance with Subarticle IWE-5240 to assess the structural condition of areas affected by repair/replacement activities. The licensee also stated that the majority of the repair/replacement activities for Category E-A components at Braidwood have been associated with the replacement of bolted components (e.g., containment personnel [air lock] components, containment penetration blind flange bolting, etc.). The licensee further stated that the existing components have been replaced either to improve reliability, because existing components were at the end of service life, or because existing material was lost or damaged during disassembly.

Class MC replacements or repairs at Braidwood have not been required due to degradation associated with the conditions described in Subarticie IWE-1241 , which would categorize components as augmented areas. The licensee also stated that, in accordance with site repair/replacement plans, when a Class MC Section XI repair or replacement is to be performed, construction code non-destructive examinations and pre-service examinations of replacement items are performed following installation of minor repairs or prior reassembly of the component, when access to surfaces is not limited, permitting the examiner full access to the existing and replacement materials requiring the pre-service inspection. The licensee stated that the appropriate time to assess the structural condition of these locations is during component disassembly. Components and connections are also inspected by mechanics as standard practice, whether or not the component is scheduled for an ASME Section XI periodic inspection.

3.1.1 Component Identification for this RR Code Class: MC Code

References:

IWE-2000 Examination and Inspection IWE-5000 System Pressure Tests

-3 Examination Categories: E-Aand E-C Item Numbers: E1.10 and E4.10

==

Description:==

IWE Components Subject to Repair/Replacement Pressure Testing 3.1.2 Applicable Code Edition and Addenda The applicable Code edition and addenda relating to this RR are the ASME Code,Section XI, 2001 Edition through the 2003 Addenda.

3.1.3 Applicable Code Requirement from which Relief is Requested ASME Code,Section XI (2001 Edition, 2003 Addenda), Subarticle IWE-S240, requires a detailed visual examination (per Subarticle IWE-2310) to be performed on areas affected by repairlreplacement activities during the post-repair/replacement pressure test required by Subarticle IWE- S220. Subarticle IWE-2310 states that detailed visual examinations shall be performed in accordance with IWE-2S00 and Table IWE-2S00-1, Examination Category E-A and E-C. The regulations at 10 CFR SO.SSa(b)(2)(ix)(G) require this detailed visual examination to be conducted using the VT-1 method.

Subarticle IWE-S221 requires all repairlreplacement activities (except those noted in Subarticle IWE-S222) performed on the pressure retaining boundary of Class MC or Class CC components to be subjected to pneumatic leakage testing in accordance with 10 CFR Part SO, Appendix J, Paragraph IV.A.

Subarticle IWE-S222 states that leakage tests for the following minor repair/replacement activities performed on the pressure retaining boundary may be deferred until the next scheduled leakage test, provided that nondestructive examination is performed in accordance with the station Repair/Replacement Program and Plan. Minor repairs include the following activities:

(a) Welds of attachments to the surface of the pressure retaining boundary (b) Weld cavities, the depth of which does not penetrate the required design wall thickness by more than 10 percent (c) Welds attaching penetrations that are NPS 1 [nominal pipe size of 1 inch] or smaller Subarticle IWE-S240 requires a detailed visual examination (per Subarticle IWE-231 0) on areas affected by repair/replacement activities that are performed during the Appendix J pneumatic leakage test.

As noted above, 10 CFR SO.SSa{b)(2)(ix)(G) requires the VT-1 method to be used to conduct the examination in Item 4.11 (detailed visual) of Table IWE-2S00-1.

-4 Additionally, 10 CFR 50.55a(b)(2)(ix)(H) requires the VT-3 method to be used to examine containment bolted connections and a subsequent VT-1 examination to be performed if any flaws or degradation are noted during the initial VT-3 examination.

3.1.4 Licensee's Basis for Request The licensee stated that performing a detailed visual examination (i.e., the VT-1 method per 10 CFR 50.55a(b)(2)(ix)(G>> for replacements or minor repairs during the Appendix J pneumatic leakage test after IWE-2200 pre-service examinations are already performed, or after the components have been re-assembled, does not provide any additional assurance of safety. The licensee also stated that the conditions of interest for the detailed visual examinations described in Subarticle IWE-1241 would be readily apparent to the examiners and mechanics while the component was disassembled. The licensee further stated that depending on the unit's Technical Specification operating mode when the minor repair or replacement is performed, plant conditions could be such that the examiner performing the detailed visual examination during the Appendix J pneumatic leakage test would have to perform the examination in a neutron radiation field (e.g., personnel [air lock] during unit operation), which would be contrary to the principles of ALARA [as low as is reasonably achievable] for occupational radiation exposure.

The licensee stated that, depending on which Appendix J surveillance is required, access to the replaced component may not be possible. For example, if an interlock barrel test is required, examiners cannot perform a meaningful examination because the area of interest cannot be entered, and access to the replaced component would be limited to looking through a sight glass (only present on the outer access door). The licensee also stated that at Braidwood, Appendix J pneumatic leakage testing is performed by operators using continuous use surveillances, along with calibrated equipment, and that performing a detailed visual examination in conjunction with the Appendix J surveillance does not provide any additional assurance of safety beyond the current Appendix J practices.

The licensee has, therefore, requested relief pursuant to 10 CFR 50.55a(a)(3)(ii), stating that compliance with the specified code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

3.1.5 Proposed Alternative Examination and Duration In lieu of performing the detailed visual examination during the Appendix J pneumatic leakage tests after a replacement or minor repair at Braidwood, the licensee proposes to perform VT-1 examinations of Class MC or Class CC component repairs/replacements along with any other required IWE-2200 examinations. The licensee stated that these examinations shall be performed in accordance with the Repair/Replacement Program prior to the conduct of the Appendix J pneumatic leakage test.

The relief is requested for the third 10-year lSI interval for Braidwood, Units 1 and 2, which are currently scheduled to end on July 28, 2018, for Unit 1, and on October 16, 2018, for Unit 2.

-5 3.2 NRC Staff Evaluation Per IWE-231 O(d), the purpose of the detailed visual examination (VT-1) required by Subarticle IWE-5240 is to assess the structural condition of areas affected by repair/replacement activities.

Subarticle IWE-5222 allows the leakage tests following the minor repairireplacement activities specifically identified in IWE-5222(a), (b) and (c) to be deferred until the next scheduled leakage test, provided that the nondestructive examination is performed in accordance with the Repair/Replacement Program and Plan.

With regards to the RR described herein, the Subarticle IWE-5240 requirements for the detailed visual examination (VT-1) during the pressure test will be completely met for the repairireplacement activities on the containment pressure boundary, except for the minor repairireplacement activities specifically identified in IWE-5222(a), (b) and (c). In lieu of the Subarticle IWE-5240 detailed visual examination (VT-1) requirement for the minor repairireplacement activities specifically identified in IWE-5222(a), (b) and (c), a non-destructive examination (such as RT, MT, etc.) of the affected area will be performed per the construction code or Repair/Replacement Program and Plan immediately following these minor repairireplacement activities for which relief is sought. Further, applicable leakage testing of the affected area will be performed in accordance with Appendix J.

The NRC staff finds that the post-repair non-destructive examination of the affected area would satisfy the intent of IWE-231 O(d) for the detailed (VT-1) visual examination required by Subarticle IWE-5240 in assessing the structural condition of the area affected by minor repair/replacement activities. Further, the applicable leakage test performed on the area affected by the minor repair/replacement, in accordance with the requirements and criteria in Appendix J would provide an acceptable measure of the structural integrity and leak tightness of the affected area. The NRC staff agrees that for minor repair/replacement activities, the VT-1 examination will not provide any additional assurance of safety beyond that provided by the nondestructive examination and leakage test. The NRC staff agrees that the presence of a VT-1 examiner in addition to the Appendix J test personnel would result in unnecessary additional dose exposure.

Based on the above, the NRC staff concludes that Braidwood, Units 1 and 2, complying with the specified requirement of Subarticle IWE-5240 would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity and leak tightness of the areas affected by minor repair/replacement activities. Therefore, the alternative proposed in RR 13R-06 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the third 10-year lSI interval for Braidwood, Units 1 and 2.

4.0 CONCLUSION

Based on the information provided in Reference 1 related to RR 13R-06, and on the NRC staff's evaluation discussed above, the alternative proposed in RR 13R-06 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the third 10-year lSI interval for Braidwood, Units 1 and 2, which are currently scheduled to end on July 28, 2018, for Unit 1, and on October 16, 2018, for Unit 2.

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

- 6

5.0 REFERENCES

1. Letter No. RS-09-118 dated September 4, 2009, from Patrick R. Simpson, EGC, to USNRC, "Submittal of Relief Requests Associated with the Third Inservice Inspection Interval" (ADAMS Accession No. ML092510267).

2. Letter dated September 22, 2008, from L. James (NRC/NRR) to R. L. Anderson (Duane Arnold Energy Center), "Duane Arnold Energy Center Safety Evaluation for Relief Requests MC-R001 and MC-P001 for Second Containment Inservice Inspection Interval" (ADAMS Accession No. ML082460235).

Principal Contributors: Dan Hoang, NRR William Jessup, NRR Date: May 11, 2010

C. Pardee -2 Please contact Mr. Marshall David at (301) 415-1547 if you have any questions on this action.

Sincerely, IRA!

Stephen J. Campbell, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. STN 50-455

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:

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