ML20127M264
| ML20127M264 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 11/25/1992 |
| From: | Burski R ENTERGY OPERATIONS, INC. |
| To: | Murley T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20127M267 | List: |
| References | |
| W3F192-0369, W3F192-369, NUDOCS 9211300106 | |
| Download: ML20127M264 (3) | |
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R. F. Durcki
,q W3F192-0369 A4.05 QA November 25, 1992 U.S. Nuclei.r Regulatory Commission ATTN:
Mr. Thomas E. Murley, Director, Office of Nuclear Reactor Regulation Washington, D.C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 10 CFR Part 50 Appendix H.III.A -
Reactor Vessel Material Surveillance Program Requirements - Report of Test Results
Dear Mr. Murley:
The subject requirement states that a summary of Reactor Vessel capsule tett results is to be submitted to the NRC within one year after capsule withdrawal from the reactor.
In letter W3F192-0052 dated April 2, 1992, Waterford 3 requested an extension for submittal of the summary for the test capsule pulled during Refuel 4 (April 11,1991).
In a letter dated June 18, 1992, the NRC granted an extension, permitting Waterford 3 to make the test summary submittal by November 30, 1992.
Enclosed is the required test report.
Wateriord 3 has revfc 7d the results of the reactor vessel surveillance specime., testing ar. concludes that there are no changes required to the pressure-temperaturd operating curves at this time. Although no changes are required at thhi time, the curves do have to be updated prior to 8 EFPY as documented in Technical Specification Figures 3.4-2 and 3.4-3.
The urveillance specimen testing vendor (B&WNS) has stated in the report (s
.on 7.4) that the current curves may be cxtended to 10.5 EFPY based ne fluence calculations performed for capsule W-97 and the results of o'
ti.e Charpy impact test results.
Based on current operations, Waterford 3 should reach 8 EFPY during cycle 7 in early 19%. Therefore, new pressure-temperature curves will be completed prior to that time.
The current withdrawal schedules shown on Table 4.4-5 of Technical Specification 4.4.8.1.2 for capsules W-83, W-104, W-263, W-277, and W-284 differ from the recommended Surveillance Capsule Removal Schedule documented in Section 9 of the test report. This difference is due te Of 9211300106 921125 PDR ADOCK 05000392 J
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USNRC 10 CFR Part 50 Appendix H.III.
- Reactor Vessel Material Surveillance Program Requirements - Report of Test Results W3F192-0369 Page 2 November 25, 1992 changes in the standards for calculating the withdrawal schedule. The original schedule, which CE provided, was developed in 1974 and was based on 10 CFR 50 Appendix H.
The schedule provided in the test report is based on the ASTM E-185-XX (To Be Released) standard and the fluence analysis. Waterford 3 will modify its withdrawal schedule. This modified schedule will change the selection of the next capsule due to the new lead factors. The original lead factors for all specimens were 1.5 but now have been more accurately determined. Two of the specimens (W-104 and W-284) have lead factors of 0.81.
These should not be used for the standard surveillance program. New specimen sample selections will be made ensuring a lead factor of greater than one is used. Additionally, the time or target fluence for the specimen pulls will be evaluated to determine the most appropriate schedule.
This decision will be based on engineering judgement including any specific regulatory 'uldance provided.
Proposed changes to the curves and schedules will be submitted to the NRC as a technical specification amendment request by December 31, 1993.
Waterford 3's surveillance requirement 4.4.8.1.2 requires the calculation of weld material E-8018 per the Regulatory Guide (RG) 1.99 method and by testing of the weld material 88114/0145.
This is insufficient to complete the analysis. A complete analysis includes the RTndt shift of the limiting material (base metal plate M-1004-2, transverse) by both measurement and by RG 1.99 analysis including considerations for limited data. Waterford 3, as part of the foregoing technical specification change amendment request, will clarify techni:a1 specification 4.4.8.1.2 to establish that the RTndt analysis should be based on the limiting material, base metal plate M-1004-2, transverse.
The testing of the limiting reactor vessel material, base metal plate H-1004-2 (transverse), yielded an RTndt adjustment of +36 F.
This is greater than the predicted increase (+18F) as calculated by the methodology of RG 1.99 Rev. 2.
As stated in RG 1.99 Rev.2, two or more credible surveillance data sets are required to use the results to determine the adjusted reference temperature.
In accordance with regulatory guide 1.99 Rev. 2, the RTndt will be calculated based on two or more credible surveillance data sets.
Additionally, the guide requires surveillance data to be evaluated by several criteria to determine the weight given this data relativt to the methodology of the guide. Specifically, the scatter in the plots of Charpy energy verses temperature should be small enough to permit the determination of the 30 ft-lb temperature " unambiguously".
The results of the testing have been plotted and evaluated by the testing laboratory and the reactor vessel manufacturer. Their independent assessments of the
4 USNRC 10 CFR Part 50 Appendix H.III.A - Reactor Vessel Material Surveillance Program Requirements - Report of Test Results W3F192-0369 Page 3 November 25, 1992 data provided a variance of at least 7 F for the delta RTndt.
The measured RTndt shift is greater than the RG 1.99 predicted shift by approximately 10 F taking into account the 7 F variance. Waterford 3 will therefore scrutinize the results of the next surveillance capsule to determine if this is a true shift greater than the Regulatory Guide prediction or an anomaly.
If you have any questions, please contact R.C. O'Quinn at (504) 739-6613.
Very truly yours, f/hwh R.F. Burski Director, Nuclear Safety RFB/0PP/ssf Enclosure cc:
J.L. Hilhoan, NRC Region IV D.L. Wigginton, NRC-NRR R.B. McGehee N.S. Reynolds i
NRC, Document Control Desk NRC Resident Inspectors Office l
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