ML20081D499
| ML20081D499 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/26/1983 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20081D502 | List: |
| References | |
| RTR-REGGD-01.150, RTR-REGGD-1.150 GL-83-15, NUDOCS 8311010220 | |
| Download: ML20081D499 (3) | |
Text
O Mailing Address
+ Allbama Power Comp:ny 600 North 18th Street Post Offica Box 2641 Birminghim. Alabam2 35291 Telephone 205 783-6081 F. L Clayton, Jr.
Senior Vice President g3hgg gg Flintndge Build:ng Docket Nos. 50-348 the southern electrc system 50-364 October 26, 1983 Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention:
Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Units 1 and 2 Generic Letter 83-15 Gentlemen:
Alabama Power Company has reviewed Generic Letter 83-15, " Implementation of Regulatory Guide 1.150 Rev. 1" and has evaluated the subject regulatory guide, " Ultrasonic Testing of Reactor Vessel Welds During Preservice and In-service Examinations." As a result, Alabama Power Company has revised its reactor vessel examination program as described herein.
Pursuant to the regulations of 10 CFR 50.55a(g), Alabama Power Company is committed to meet the 1974 Edition through the Summer 1975 Addenda of the ASME Code Section XI requirements for the inservice inspection of Farley Nuclear Plant Units 1 and 2.
These requirements include the Mandatory Appendix I requirements entitled " Ultrasonic Examination," which govern reactor vessel examination. While these requirements remain as the technical basis of the examination program, Alabama Power Company has developed an Augmented Reactor Vessel Examination Program for both Farley Nuclear Plant units. A copy of this program is enclosed for your review. This augmented program will provide an acceptable alternative method for implementing the guidelines specified in Appendix A of Regulatory Guide 1.150 Rev. 1.
This program is intended to be implemented for the remainder of the first ten year inspection interval for each unit. For Unit 1, the program will be implemented during the 80 and 120 month examinations (the 40 month examination was completed prior to July 15, 1981) and during the 40, 80 and 120 month examinations for Unit 2.
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i Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Units 1 and 2 Generic Letter 83-15 October 26, 1983 Page 2 Alabama Power Company's program utilizes portions of the Ad Hoc Committee recommendations adopted by the NRC in Appendix A of Regulatory Guide 1.150 Rev. 1.
A Technical Position is included in the enclosed program which describes the inspection system performance checks, equipment calibration, near surface resolution, beam profile, scanning weld-metal interface, record-ing and sizing of indications and reporting requirements. The 20 per cent DAC recording levels will be utilized as recommended in Appendix A for the reactor vessel beltline region weld examinations. In accordance with the provisions stipulated in Alabama Power Company's Augmented Reactor Vessel Examination Program, exception is taken to the recording levels recommended in Appendix A for the non-beltline region reactor vessel welds. Reflectors in the full wall thickness of the non-beltline region reactor vessel welds will be recorded at 50 per cent DAC in accordance with the ASME Code requirements.
This approach is consistent with the Pressurized Thermal Shock (PTS) analysis. The results of this analysis indicate that the beltline region is the controlling area of concern for postulated vessel failure. Therefore, inspection and evaluation of the beltline region welds in accordance with Alabama Power Company's program will provide an acceptable basis of reactor vessel integrity.
Both vessels were analyzed as a part of the Westinghouse Owners Group Pressurized Thermal Shock (PTS) review. The Westinghouse Owners Group has developed, and submitted for NRC consideration, conservative screening criteria to identify reactor vessel susceptibility to PTS. Using this screen-ing criteria, Westinghouse's evaluation of the Farley reactor vessels described in WCAP-10019, " Summary Report on Reactor Vessel Integrity for West-inghouse Operating Plants" indicates that the Farley vessels are not expected to reach the screening criteria before the end of plant life.
Alabama Power Company has taken additional measures to reduce the effects postulated in the PTS analysis by utilizing a revised fuel matrix to lower the neutron flux leakage from the core. In addition, specific plant procedures, including training of personnel, are being implemented to address the PTS concern. These procedures are being implemented in accordance with the NUREG-0737, Supplement 1 commitments described in Alabama Power Company's letter of April 15, 1983 to the NRC.
Therefore, utilization of the Appendix A recommended recording levels only in the beltline region welds will provide adequate assurance of vessel integrity in the controlling beltline region.
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I Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Units 1 and 2 Generic Letter 83-15 October 26, 1983 Page 3 The application of the 20 per cent DAC recording levels recommended in Appendix A to the less critical non-beltline region welds may result in the extension of a planned outage due to the unnecessary recording, sizing and evaluation of insignificant indications. Industry practice has shown that the recording of examination data at 20 per cent DAC will produce a large number of indications which, upon evaluation, will probably be found acceptable and would increase the length of a refueling outage with no measurable concomitant increased confidence in the vessel's integrity. It is the judgement of Alabama Power Company that significant indications recorded it 20 per cent DAC will also be recorded at 50 per cent DAC. In addition, unknown cladding surface conditions may produce noise interference which could complicate
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indication sizing and recording and may even prevent evaluation of data recorded at 20 per cent DAC. Alabama Power Company's program will provide,an effective method for conducting ultrasonic examinations of reactor vessel welds, while minimizing the number of insignificant indications requiring evaluation. This will reduce the significant critical path outage time which could be required for the recording, sizing and evaluation of meaningless data.
Alabama Power Company has reviewed the Technical Specification require-ments related to the ultrasonic examination of reactor vessel welds. Since the Technical Specifications reference the ASME Code and do not delineate specific reactor vessel examination requirements (other than the ASME Code), a Technical Specification change is not required.
Yours truly, f,
3 F. L. Clayton, Jr.
STB:kc/D-317 Enclosure cc:
Mr. R. A. Thomas Mr. G. F. Trowbridge Mr. J. 2. "'Reilly Mr. c. A. Reeves Mr. W. H. Bradford i
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