ML20059J577
| ML20059J577 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/05/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059J561 | List: |
| References | |
| NUDOCS 9311120270 | |
| Download: ML20059J577 (3) | |
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wAsninoTou. o.c. 20sss-oooi SAFETY EVALUATION BY THE OFFICE NUCLEAR REACTOR REGULATION OF THE FIRST 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN FOR NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION. UNIT 3 DOCKET N0. 50-423
1.0 INTRODUCTION
The Technical Specifications for Millstone Nuclear Power Station, Unit 3 state that the inservice inspection and testing'of American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and t
applicable Addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to i
The Code of Federal Regulations at 10 CFR 50.55a(a)(3) state that alternatives to the requirements of paragraph
_t (g) may be used, when authorized by the NRC, if (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 l
difficulties without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design,
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geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the second 10-year interval comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120 month inspection interval, subject to the limitations and modifications listed therein.
The applicable edition of Section XI of the ASME Code for Millstone Nuclear Power Station, Unit 3 First 10-Year Inservice Inspection (ISI) Interval is the 1983 Edition, through Summer 1983 Addenda. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein.
The ASME Code cases that the staff has determined suitable for use are listed in NRC Regulatory Guide 1.147 " Inservice Inspection Code Case Acceptability-ASME Section XI Division 1."
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Pursuant to 10 CFR 50.55a(g)(5)(iii), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME Code requirement.
After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
In a letter dated May 26, 1993, Northeast Energy Company (the licensee) submitted its resolutions regarding the staff's concerns in NRC's Safety Evaluation (SE) dated March 3, 1993, on Millstone Nuclear Power Station, Unit 3 First 10-Year Interval Inservice Inspection Program Summary Report, Rev. 3.
In addition, the licensee withdrew Requests for Relief IR-15 and IR-17, and requested approval to use ASME Code Case N-491 " Alternative Rules for Examination of Class 1,2,3 and MC Component Supports of Light-Water Cooled Power Plants,Section XI, Division 1" and Code Case N-307-1 " Revised Ultrasonic Examination Volume for Class 1 Bolting, Table IWB-2500-1, Examination Category B-G-1, When the examinations Are Conducted from the Center-Drilled Hole,Section XI, Division 1" as alternatives to the Code recuirements.
2.0 EVALUATION The staff has evaluated the information provided by the licensee in support of the licensee's resolutions of three staff items of concerns in the NRC SE dated March 3, 1993, and request to invoke ASME Code Cases N-491 " Alternative Rules for Examination of Class 1,2,3 and MC Component Supports of Light-Water Cooled Power Plants,Section XI, Division 1" and N-307-1 " Revised Ultrasonic Examination Volume for Class 1 Bolting, Table IWB-2500-1, Examination Category B-G-1, When the examinations Are Conducted from the Center-Drilled Hole,Section XI, Division 1."
i A.
The first staff concern involved the exclusion of the chemical and volume 1
control system and the high pressure safety injection system from having a volumetric examination performed on 7.5 percent of welds in these systems.
The licensee has proposed to include augmented volumetric examination of 7.5 percent of welds in the above systems.
The staff has found the licensee's proposed augmented volumetric examinations acceptable and that the Hi11 stone Nuclear Power Station, Unit 3, First 10-Year Interval ISI i
Program Summary Report, Rev. 3 is in compliance with the regulations.
I a
.- B.
The second and third staff concerns referred to Requests for Relief IR-15 and IR-17 which were denied by the staff in its SE dated March 3, 1993.
In its letter dated May 26, 1993, the licensee has proposed to withdraw both requests for relief and to invoke ASME Code Cases N-491 and N-307-1.
Code Case N-491 defines new requirements for selection of component supports including examination of at least 10 percent of Class 3 piping supports, and 100 percent of supports other than piping supports.
Based on the f act that Code Case N-491 has been approved in Regulatory Guide (RG) 1.147, Revision 10, dated July 1993, the staff determined that the licensee's use of the Code Case is acceptable.
Code Case N-307-1 pertains to a revised ultrasonic examination volume for Class 1 Bolting.
The licensee proposes to apply Code Case N-307-1 to Class 1 Reactor Vessel Studs and Class 2 Main Steam Isolation Valve Studs.
The use of Code Case N-307-1 is only acceptable for Class 1 Bolting, Table IWB-2500-1, Examination Category B-G-1 and does not apply to Class 2 Bolting. Therefore, the licensee cannot apply Code Case N-307-1 to the Class 2 Main Steam Isolation Valve Studs.
Based on the fact that Code Case N-307-1 has been approved by the NRC fn RG 1.147, Revision 10, the staff finds it acceptable for the licensee to apply Code Case N-307-1 to -
Class 1 Reactor Vessel Studs that are in Examination Category B-G-1,
3.0 CONCLUSION
The staff has reviewed and evaluated the licensee's resolutions to the staff concerns contained in the NRC SE dated March 3, 1993, and has concluded: (1) that the licensee's resolutions are acceptable, and (2) that the #177 stone Nuclear Power Station, Unit 3, First 10-Year Interval ISI Program Summary Report, Rev. 3 is in compliance with the regulations.
Code Cases N-491 and N-307-1 have been determined to be acceptable for use and are listed in RG 1.147, Revision 10, dated July 1993, as such. The staff has concluded that the licensee may invoke both Code Cases for use, but should note that Code Case N-307-1 does not apply to Class 2 Bolting and, therefore, may not be used for the Class 2 main steam isolation valve studs.
Principal Contributor:
T. McLellan Date: November 5, 1993 1
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