ML20116A927
| ML20116A927 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/01/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20116A926 | List: |
| References | |
| NUDOCS 9210300213 | |
| Download: ML20116A927 (3) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION y
g WASWNGiot D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 94 10 FACILITY OPERATING LICENSE NO. NPF-2 SOUTHERN NUCLEAR OPERATING COMPANY._lNC.
JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1 DOCKET NO. 50-348 1.0 1"TRODUCT10N By letter dated October 29, 1991, as supplemented July 1, 1992, the Southern Nuclear Operating Company, Inc. (the licensee), submitted a request for changes to Joseph M. Farley Nuclear Plant (Farley), Unit 1, Technical Specification (TS) Section 3.4.7.2 and Bases Sections 3/4.4.6 and 3/4.4.7.2.
The requested changes would reduce the steam generator (SG) primary-to-secondary leakage limits.
The change lowers the primary-to-secondary leakage limit through all SGs from one gallon per minute (gpm) (1440 gallons per day (gpd)) to 420 gpd, and lowers the primary-to-secondary leakage limit through any one SG from 500 gpd to 140 gpd.
The July 1, 1992, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
2.0 EVALUATION Inservice inspections of the steam generators at Farley, Unit 1, have revealed the presence of circumferentially oriented defects in the steam generator tubing.
The licensee has proposed a change to its TS to adopt a more restrictive limit for primary-to-secondary leakage based on the discovery of these circumferentiall,v oriented defects.
The NRC staff reviewed the licensee's proposed changes ~and subsequently requested additional information from the licensee by letter dated June 3, 1992.
The licensee responded to this request by letter dated July 1, 1992.
The current 500 gpd primary-to-secondary leakage limit per SG is intended to assure that cracks which leak :.t a rate up to this limit will have an adequate margin of safety, consistent with the guidance in Regulatory Guide 1.121,
" Bases for Plugging Degraded PWR Steam Generator Tubes," to withstand the load imposed during normal operation and by postulated accidents.
The current one gpm primary-to-secondary leakage limit through all SGs is consistent with the assumptions used in the Farley, Units 1 and 2, Final Safety Analysis Report (FSAR) design basis accident analyses.
The circumferential defects are primarily located within the WEXTEX expansion transition area of tubes located in the central, sludge pile area and'also in the tangent point and apex locations of the U-bend region of row I tubes.
Several of the indications that were found could be traced back, using hindsight, to distorted indications in previous outages; however, a growth 9210300213 921001 PDR ADOCK 05000348 p
J i I rate could not be calculated based on the previous inspection results because l
the earlier bobbin coil inspection data could not be directly correlated to the new rotating pancake coil probe data. Tubes identified with circumferential indications were removed from service by plugging.
Row 1 U-bend tubes, in which circumferential cracking was not detected, were heat treated during the last outage to minimize any residual tensile stresses in the tubing and to minimizc crack growth in subsequent plant operation.
The licensee has calculated a projected end-of-cyc'.e circumferential crack angle using an estimate for the non-destructive examination (NDL) detection threshold, corrosion growth allowance, and NDE uncertainty.
The maximum projected end-of-cycle circumferential crack angle is expected to be within the maximum allowable single, through-wall circumferential crack angle for meeting Regulatory Guide 1.121 guidance for normal, upset, and accident loading conditions with respect to tube burst.
If the cracking were to continue at Farley, Unit 1, the licensee expects the development of a segmented crack morphology around tne tube circumference with through-wall cracks of various arc lengths separated by ligaments.
This expectation is based on pulled tube evidence from several plants with WEXTEX expansions.
The leakage from this expected segmented crack morphology (approximately 35 degree cracks separated by 20 mil ligaments) would exceed the proposed 140 gpd per SG primary-to-secondary leakage limit.
Similarly, the expected leakage if two of these individual cracks were to grow together (i.e. loss of a ligament) would exceed the 140 gpd limit, Therefore, the licensee concluded that the 140 gpd limit provides assurance that should a circumferential crack propagate at an unexpectedly high rate or if the ligaments separating individual cracks should rupture, sufficient time would exist to shutdown Farley, Unit 1, prior to a SG tube rupture.
The licensee is capable of detecting operational leakage of this magnitude with existing radiation monitors.
3.0
SUMMARY
The licensee has determined that the proposed 140 gpd primary-to-secondary leakage limit is significantly less than the leakage expected due to the maximum allowable single, through-wall circumferential crack angle which meets Regulatory Guide 1.121 guidance for normal, upset, and accident leading conditions with respect to tube burst. Additionally, the proposed more restrictive primary-to-secondary leakage limit will require unit shutdown at a lower primary-to-secondary leakage level and provides added assurance that the leak rate limit under normal operation will be exceeded prior to exceeding the largest permissible crack. Therefore, the staff concludes that the proposed primary-to-secondary leakage limits should provide adequate assurance of structural and leakage integrity of the steam generator tubing at Farley, Unit 1.
4.0 STATE CONSULTATI.03 In accordance with the Commission's regulations, the State of Alabama of ficial was notified of the proposed issuance of the amendment.
The State official had no comments.
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5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any ffluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideratin, and there has been no public comment on such finding (57 FR 2580).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
K. Karwoski Date: October 1, 1992 l
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