ML20210P832
| ML20210P832 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 08/11/1999 |
| From: | Hutton J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9908130090 | |
| Download: ML20210P832 (6) | |
Text
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PECO NUCLEAR
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A Unit of PECO Energy
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August 11,1999 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR 56 U. S. Nuclear Regulatory Commiss!on Attn: Document Control Desk Washington, DC 20555 i
Subject:
Peach Bottom Atomic Power Station, Units 2 and 3 Submittal of Proposed Altematives to the Requirements of 10 CFR 50.55a(g)(6)(ii)(B)(1) Conceming the Containment Inservice inspection Program
Reference:
Let^er from G. D. Edwards (PECO Energy Company) to U. S. Nuclear Regulatory Commission, dated February 17,1999
Dear Sir / Madam:
In the Referenced letter, PECO Energy Company submitted 11 propoC attematives to the requirements of 10 CFR 50.55a(g)(6)(C)(B)(1), conceming the implementation of a Containment inservice Inspection Program and completion of certain examinations by September 9,2001. Attachment 1 contains information in response to a telephone conversation between PECO Energy Company and U. S. Nuclear Regulatory Commission staff on Thursday, July 15,1999, regarding some of the subject proposed attematives.
If you have any questions, please contact us.
Very truly yours, O1]
I.
James A. Hutton r
Director-Licensing Attachment i
cc:
H. J. Miller, Administrator, Region I, USNRC A. C. McMutray, USNRC Senior Resident inspector, PBAPS i
9908130090 990811 PDR ADOCK G5000277 PDR:
O L
1 ATTACHMENT 1 ADDITIONAL INFORMATION REGARDING PROPOSED ALTERNATIVES TO THE REQUIREMENTS OF 10 CFR 50.55a(gM8Mii)(B)(1) CONCERNING THE CONTAINMENT INSERVICE INSPECTION PROGRAM Requests CRR-01 through CRR-11 i
i
i August 11,1999 fage.1, Question:
- 1. Relief Request CRR-02, " Alternative Requirements for Qualification and Certification of Nondestructive Examination Personnel" It is not clear to the staff whether this relief request is to be applied to the current ten year interval or to all future inspection intervals.
Response
As discussed in the " Applicable Time Period" section of Relief Request CRR-02, this relief is requested for the first ten-year containment inspection interval of the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Inservice
" Inspection Program. This program began November 5,1998, and will conclude November 4,2008, at which time a future version of the ASME XI Code is expected to be adopted, in accordance with current regulations.
Question:
- 2. Relief Request CRR-03," Alternative Criteria for Preservice of Reapplied Paint and Coatings," and CRR-04, " Alternative. Criteria for Visual Examination of Paint or Coating Prior to Removal" in the basis for attemative section for these two relief requests, the licensee states that the PECO Energy Coatings Program is to be used for examining and verifying the adequacy of coatings, and will provide adequate level of safety and quality. With respect to this program, the licensee needs to provide the following information related to the protective coating used inside the containment:
A. Standards, procedures or guidelines used for (a) surface preparation, application, surveillance, and maintenance activities for protective coatings, and (b) documenting the quality and condition of reapplied paint.
B. Maintenance activities including reworking of degraded coating, removing of
~ degraded coating to soundcoating, preparing the surface, applying new coating, and verifying the coating quality.
Response-The coatings on the interior surface of the containment vessel are considered safety-related. They are applied and inspected in accordance with the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3,10 CFR 50, Appendix B Quality Assurance Program. This program is described in Section D.11 of the
August 11,1999
.Page 2, PBAPS, Units 2 and 3 Updated Safety Analysis Report (USFAR). Appendix 17.2A.8 of the Appendix B Quality Assurance Program endorses ASTM D3843-93 (" Standard Practice for Quality Assurance for Protective Coatings Applied to Nuclear Facilities"). ASTM D3843-93 replaces Regulatory Guide 1.54, " Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants," and ANSI Standard N101.4-1972, " Quality Assurance for Protective Coatings Applied to Nuclear Facilities." The following requirements are applicable for coatings applied to the interior surface of the containment vessel:
The quality assurance requirements of Sections 6 and 7 of ASTM D3843-93 applicable to the coating manufacturer are imposed on the coating manufacturer through the procurement process.
Coating application procedures are developed based on the manufacturer's recommendations for application of the selected coating systems.
Coating applicators are qualified to demonstrate their ability to satisfactorily apply the coatings in accordance with the manufacturer's recommendations.
Quality Verification (QV) personnel qualified in accordance with ANSI o
N45.2.6-1978 and ASTM D 4537-91 perform inspections to verify conformance to the coating application procedures and perform examinations of previously applied coatings. Section 10 of ASTM D3843-93 is used as a guideline in the establishment of the inspection program.
Alternatively, personnel qualified in accordance with ASNT SNT-TC-1 A (8/84) supplemented with appropriate coatings training, perform examinations of previously applied coatings.
Documentation demonstrating conformance to the above is maintained.
A visual coatings examination of accessible and immersed surfaces of the drywell and suppression / torus surfaces is performed at least every four (4) to six (6) years in accordance with 10 CFR 50.65, " Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." These periodic examinations will identify evidence of flaking, blistering, peeling, discoloration, or other signs of coatin0 distress which might be indicative of degradation of the containment structuralintegrity.
This program is prerently being evaluated against the recommendations of Information Notice (IN) 97-13, " Deficient Conditions Associated with Protective
- Coatings at Nuclear Power Pirents." and the Electric Power Research Institute (EPRI) Plant Support Engineering TR-109937, " Guideline on Nuclear Safety-
August 11,1999
.Rege 3, Related Coatings". Changes to improve this program will be implemented as necessary.
Question-
- 3. Relief Request CRR-05, " Alternative Criteria for VT-2 Visual Examination Following Repair, Replacement or Modification" in the proposed alternative criteria section, the licensee states that in addition, examinations following repairs or replacements on containment components will be performed in accordance with PECO Energy Repair / Replacement Specification M-679. The staff would like to know some details about this program.
Response-Specification M-679 concerns the ASME Section XI Repair and Replacement Program. This program provides the administrative guidance for satisfying the requirements of the Section XI Code, as applicable to repairs and replacements of Class 1,2,3, and MC components and their supports. Sections 8,9, and 10 of this specification address the need for satisfying the construction code requirements, Section 12 addresses pressure testing of components following repair or replacement, and Section 13 addresses Preservice Inspections (PSI) required of repaired or replaced components. Accordingly, a repair of a Class MU component would be implemented (as required by the repair and replacement program) in accordance with the rules of the original construction code. After completion of the repair / replacement, the original construction code-required NDE would be performed. Following this, the ASME Section XI requirement for Preservice Inspection (PSI) would be performed, in accordance with the Containment ISI Prngram. As required by Section XI, the method of inspection for the PSI would be the method originally used to detect the
' condition which requ: red the repair / replacement, and/or the method required for subsequent inservice Inspections (ISI). In the case of Class MC components, this method would normally be the visual (VT-3) method. If the repaired or replaced area were Category E-C, Containment Surfaces Requiring Augmented Examination, then tne method of inspection for the PSI would be the Visual VT-1 method or Ultrasonic Testing (UT). These examinations will confirm the structural integrity of the repaired or replaced area of the containment.
Confirmation of the leak-tight integrity of the area will then be verified by a pressure test. The system pressure testing would be conducted, as applichble, in tha area of the repair or replacement, per 10 CFR 50, Appendix J. The pressure testing would be conducted by personnel trained in the methods of testing the containment vessel, as required by Appendix J, utilizing equipment and procedures routinely used for the periodic pressure testing of the contawiment. Performance of the visual VT-2 examination, during the conduct of i
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I this pressure test, would in most cases be impractical, due to accessibility.
Access to perform the visual examination of the repaired / replaced area is normally prohibited by either encapsulation of the pressure test boundary (i.e.
Local Leak Rate Test) or personnel access restrictions into containment during testing (Integrated Leak Rate Test). VT-2 exammation of the repaired / replaced i
area from the outside surface of the containment (during the pressure test) would be meaningful and practical in some cases. However, many portions of the containment me inaccessible on the outside surface. Further,10CFR50, Appendix J acceptance criteria for the results of the pressure testing assures that the leak tight integrity of the containment vessel will support NRC safety goals.
The above described examinations and testing assure that the stnJctural integrity and leak-tight integrity of the primary containment will be maintained following any repairs or replacements of the pressure boundary. Nevertheless, a VT-2 visual examination will be performed from the outside surface of the containment, whenever access from the outside surface is available in the area of the repair or replacement being pressure tested.
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