ML17219A464
| ML17219A464 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 03/16/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17219A462 | List: |
| References | |
| NUDOCS 8703260328 | |
| Download: ML17219A464 (4) | |
Text
~g AKCy Cy I
nO I
c IA
/y Wp*y4 UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FLORIDA POWER & LIGHT COMPANY ST.
LUCIE PLANT UNIT NO.
1 DOCKET NO. 50"335
.REANALYSIS OF POSTULATED LOSS OF NON-EMERGENCY AC POWER AND SEIZED ROTOR EVENTS
1.0 INTRODUCTION
In the Safety Evaluation for St.
Lucie, Unit 1, License Amendment No.
48 (Reference 1), which authorized an increase in the licensed power from 2560 to 2700 MWt, it was found that three items needed further attention.
One of these items was closed out via Reference';
two items remained open.
The first open item was an analysis ef the loss of non-vital AC power, taking into consideration the single failure criterion.
The second open item was an analysis of the seized reactor coolant pump rotor event, taking into consideration loss of offsite power and the single failure criterion.
Florida Power & Light Company (FP8L) reanalyzed these two postulated events and submitted the results to the NRC on August 31, 1982.
Following requests for additional information, FP&L made second submittals on November 21 and December 22 of 1983, a third on September 11, 1984 (Reference 3),
and a
fourth on January 20, 1987 (Reference 4).
- 2. 0 EVALUATION 1.
Loss of All Non-Emergency AC Power This event is initiated by a loss of all non-emergency AC power after which the reactor coolant flow starts to decrease.
The reactor is tripped by a low coolant flow rate signal, which is set to actuate at or above 93 percent of the initial flow.
For the first few seconds of the transient, the Loss of Non-Emergency AC Power event behaves like a complete Loss of Forced Reactor Coolant Flow event.
The minimum departure from nuclear boiling ratio (DNBR) limit was not exceeded and no fuel failure was predicted for that event.
Thus, the DNBR limit will not be exceeded for the Loss of Non-Emergency AC Power event.
The worst single failure (WSF) for this event was found to be a stuck open Atmospheric Dump Valve (ADV).
An ADV could stick open when the pressure in the secondary system increases due to the loss of the con-denser on the loss of non-emergency AC power.
The failure of the diesel generator'to start was also considered as a candidate for the WSF but was rejected since no equipment supplied with power by the diesel (which would then be unavailable if it failed) increases the radiological dose as significantly as failure of the ADV.
Over-pressure is not a concern from a loss of the diesel since overpressure 7
P
~
I W
LA
~
P
'0 I%I
-6 411 g<<(
)
~
I 144 6 kl VL IE4 <<4 i4JC =Sdk Sg Vgl 1 Ak
)
- ~ 0,
~
d 4
criteria have been met by both St.
Lucie Unit 1 and St.
Lucie Unit 2 for complete station blackout (including diesel generator failures) as part of licensing actions for Unit 2.
Failure of an ADV in the open position maximizes steam release from the secondary side and therefore also maximizes radiological dose.
The staff agrees that an AOV sticking open during this event is the worst single failure.
For this analysis, one of Unit 1's two ADVs was assumed to fail in the open position at t=0.
This failure is a postulated initial condition chosen to maximize the potential for fuel damage and radiological release and is not a predicted consequence of the progression of the events during the transient.
This assumption conservatively envelopes the effects of either the failure of an automatically operated ADV
- system, an operator error, or a mechanical failure since the excessive cooldown starts immediately.
(An automatic
- system, actuated by an increasing secondary pressure would not function I.or fail] until later in the transient.)
The second AOV was allowed to function in accordance with an assumed automatic program.
This provides for earlier and greater cooling than would be obtained by using the Rain Steam Safety Valves (MSSV's) because the HSSV's have a higher opening setpoint.
The staff agrees that these are conservative assumptions.
The operators were assumed to begin controlling the plant cooldown at 30 minutes, using the operable ADV and the auxiliary feedwater, and closing the ADV block valve to isolate the stuck open AOV.
The staff agrees that this is a conservative assumption.
The total amount of steam released to the environment through the ADV's and HSSV's at the time the Shutdown Cooling System is started was calculated to be 913,780 lbs.
With the Technical Specification primary to secondary leak rate of 1 GPM and the maximum concentrations of radioactivity allowed by the Technical Specifications in the reactor coolant system and steam generators, the calculated 0-2 hour site boundary doses are:
Thyroid (DEq I-131)----1. 53 REH Whole Body (OEq Xe-133)--.0009 REH These are a very small fraction of the 10 CFR Part 100 limits and are acceptable.
2.
Seized Rotor Accident with Loss of Offsite Power This event is postulated to be initiated by the mechanical seizure of one of the four reactor coolant pump shafts.
Thus, there is a rapid decrease in the reactor coolant flow.
The reactor and turbine are tripped by a low coolant flow rate signal, which is set to actuate at or above 93 percent of the initial flow.
Since at full power Unit 1 produces less than SX of the capacity of the Florida electrical grid, the effect of its turbine trip on the grid
is not immediate.
The licensee
'has shown that offsite power at Unit 1 would not be lost for at least 3 seconds after the turbine trip.
Since this is a rapid transient in which the minimum ONBR is cal-culated to occur at 1.7 seconds after the seizure, the lost of offsite power has little effect on the number of fuel pin failures.
The staff agrees with this conclusion.
To maximize the offsite dose for this event, it was assumed that one of the two AOV's stuck open at the same time as the shaft seizure.
The operators were assumed to begin controlling the plant cooldown at 30 minutes after the seizure, using the operable ADV and closing the AOV block valve to isolate the stuck open ADV.
The staff agrees that this is a conservative assumption.
The total amount of steam released to the environment through the ADV's and HSSV's at the time the Shutdown Cooling System is started was calculated to be 903,094 lbs.
With the 1.63X fuel Pin failures that were calculated to occur at the minimum DNBR, the predicted 0-2 hour site boundary doses are:
Thyroid (OEq I-131)36.1 REH Whole Body (DEq Xe-133)--.06 REH Since these are within the 10 CFR Part 100 guidelines they are acceptable.
- 3. 0 CONCLUSION The staff finds the licensee's analyses of the postulated loss of non-emergency AC power and seized rotor events, as documented in submittals to the NRC dated August 31, 1982; November 21, 1983; December 22, 1983';
September ll, 1984; and January 20,. 1987, are acceptable.
4.0 REFERENCES
1.
Letter from C.
C. Nelson, USNRC, to R.
E. Uhrig, Florida Power 8 Light
- Company, November 23, 1981.
2.
Letter from R.
A. Clark, USNRC, to R.
E. Uhrig, Florida Power 8 Light Company, April 26, 1982.
3.
Letter from J.
W. WIlliams, Jr., Florida Power 8 Light Company, to J.
R. Hiller, USNRC, September 11, 1984.
4.
Letter from C. 0.
Woody, Florida Power 8 Light Company, to USNRC, January 29, 1987.
Date:
Harch 16, 1987 Principal Contributor:
E.
Lantz