ML19324C231
| ML19324C231 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 11/09/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19324C230 | List: |
| References | |
| NUDOCS 8911150283 | |
| Download: ML19324C231 (2) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO SAFETY EVALUATION LINE BREAK ALABAMA POWER AND LIGHT COMPANY JOSEPH H. FARLEY NUCLFAR PLANT, UNITS 1 AND 2
_OCKET N05. 50-348 AND 50-364 D
1.0 INTRODUCTION
In December 1987, Joseph M. Farley Nuclear Plant, Unit 2, (Farley 2) experienced a crack in a safety injection line attached to the reactor coolant system (PCS) cold leg. As part of subsequent staff reviews of this event, the staff inquired whether a break of the safety injection line was analyzed as part of the Farley Units 1 and 2 plant small break loss of coolant'acciaent (LOCA) analyses. Of specific concern was the assumption regarding safety injection _
flow in the analysis.
Unlike other cold leg breaks, a break of the safety injection line results in direct spillage of the safety injection flew to the containment and a decreased fraction of the injection flow is delivered to the intact cold legs.
In a letter dated January 14, 1988, the licensee responded to staff's concerns regarding a break in the safety injection line. The staff's evaluation of the licensee'r rehense follows.
2.0 EVALUATION In its letter, the licensee stated that the small break LOCA evaluation for Farley, Units 1 and 2, were performed for breaks in the cold leg piping. The review of the analysis showed that the broken loop safety injection flow was dssumed to spill to RCS backpressure, not Containment backpressure. Ihus, the l
licensee concluded that the small break LOCA for the Final Safety Analysis Report (FSAR) did not bound a doube-ended severance of the safety injection line.
L An evaluation was performed by the licensee to assess the etfect of spilling L
the broten loop safety injection flow to containment backpressure for a double-ended severance of the safety injection line.
Since the diameter of 1
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the saf ety injection line is 6 inches, the licensee used the small break LOCA analysis results for a 6-inch equivalent diameter break at the bottom of the cold leg.
It should be noted that the 6-inch small break is also the limiting small break LOCA for farley.
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The licensee estimated the ef fect of the decreased saf ety injection flow on the start of core uncovery and the additional time it would take to recover I
the core. These estimates assumed that the overall RCS behavior woulo not be i
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significantly affected by the decreased safety injection flow, Using the maximum core heating rate for the 6-inch line break, the licensee estimated that the additional core uncovery time (about 4 seconds) would result in an increased peak cladding temperature of 46'F.
Dased upon its assessment, the licensee concluded that the peak cladding temperature for the safety injection line break would be 1875'F for Unit I and 1758'F for Unit 2.
These temperatures are less than the 2200*F criteria of 10 CFR 50.46 and are also bounded by the calculated 2013*F peak cladaing temperatures for a large break LOCA for Farley, Units 1 and 2.
The staff has reviewed the licensee's estimates ano believes that the methoos employed and conclusions reached are reasonable. As a result, the staff concludes that Farley, Units 1 and 2, sati:,ty the performance requirements of 10 CFR 50.46 for a safety injection line break, Section 1.C.1 of Appendix K requires that a break spectrum analysis be performed to determine the break size and location which result in the maximum cladcing temperature.
In the Farley FSAR, both large break and small break spectrum evaluations, using different LOCA evaluation models, are reported. For the small break LOCA analysis, the Farley FSAR shows the 6-inch line break to be the worst case break.
However, as noted above, the licensee now estimates that the safety injection line break results in a hisher peak cledding temperature. The staff concludes that failure to properly model the safety injection lire break constitutes an error in the application of an acceptable LOCA evaluation model to Farley, Units 1 and 2.
However, since the estimated effect of this error is less than 50*F, it is not a significer.t change in the calculateo peak fuel cladding temperature as defined in 10 CFR 50.46(3)(1).
Thus, no calculations are required.
Since this is an increase in the limiting small break LOCA transient, the licensee shoulc monitor this error, and any subsequent changes and errors, and report them annually to the Commission as spec 1fied in 10 CFR 50.46(3)(.1).
3.0 CONCLUSION
S The staff concludes that the licensee's evaluations demonstrate that Farley, Units 1 and 2, satisfy the performance requirements of 10 CFR 50.46.
Thus, the staff concludes that continued operation of Farley, Units 1 and 2, pose no uncue risk to the public health and safety.
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Principal Contributor:
R. Jones I
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Dated: November 9, 1989 l
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