ML13308B935
| ML13308B935 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 08/18/1981 |
| From: | Moody W Southern California Edison Co |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-15-04, TASK-15-4, TASK-RR NUDOCS 8108240068 | |
| Download: ML13308B935 (23) | |
Text
REGULATOR OmNFORMATION DISTRIBUTION Si*EM (RIDS)
ACCESSIONJ N3R:8108240065 DOC.DATE: 81/08/18 NOTAqI7ED: NO
)0cKET #
FACIL:50-206 San Onofre Nuclear Station, Unit 1, Southern Califort 05000206 50-361 San Onofre Nuclear Station, Unit 2, South'ern Califor 05000361 50-362 San Onotre Nuclear Station, Unit 3, Southern Califor!
AUTH.NAME-AUTHOR AFFILIATION
,m000Y,
,C, Southern California Edison Co.
RECIPvAM'E RECIPICNf AFFILIATION CRUTCHFIEL),D.
Ooerating Reactors 9ranch 5 SUSJECT: Forwards draft dssessment of SEP Topic XV-4, "Loss of
,Nonemergency AC Power to Station Auxiliaries."
DISTRIBJTION CODEI A035S COPIES RECEIVED:LTR -1 ENCL L SIZE:....
TITLE: SEP Tooics NOTE3:1 cy each:SEP Section Leader & J Hanchett (Region V) 05000206 Send all FS4R s :P amends to L Chandler, 05000361 1 cy:J Hanchett (Region i),D Scalettil cy of all envir info Send all FSAR & ER amends to L Chandler.
05000362 1 cy:J Hanchett (Region V).D Scalettil cy of all envir info RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAMEi LTTR ENCL ID CODE/NAME LTTR ENCL ACTION:
ORS 45 BC 04 7
7 LIC BR #3 BC 014 7
7 INTERNAL; A/D 4ATLLOUAL15 1
1 CONT SYS A 07 1
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447 1941 47 O7 TOTAL NUmBER OF CDPIES REUIRED: LTTR ENCL
Southern California Edison Company P. C. BoX 800 2244 WALNUT GROVE AVENUE ROSEMEAD CALIFORNIA 91770 W. C. MOODY TELEPHONES MANAGER, NUCLEAR ICENSING August 18, 1981 213 2-117 1213 572-laO6 Director, Office of Nuclear Reactor Regulation Attention:
D. M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Gentlemen:
Subject:
SEP Topic XV-4 San Onofre Nuclear Generating Station Unit 1 Enclosed is the draft assessment for SEP Topic XV-4, Loss of Non-Emergency AC Power to the Station Auxiliaries. If you have any questions on this draft topic assessment or require additional information, please let me know.
Very truly yours, Enclosure 8108240068 80818 PDR AD0CK 05000206 P
0 e
XV-4 LOSS OF NON-EMERGENCY AC POWER TO THE STATION AUXILIARIES I. INTRODUCTION Loss of non-emergency ac power to the station auxiliaries can result from a failure of transmission lines or a station transformer or from a malfunction in the onsite ac distribution system. The event will trip all the reactor coolant pumps simultaneously causing a flow coastdown and a decrease in heat removal by the secondary system. In general, redundant equipment, connections, and buses are provided to minimize adverse effects so that power can be supplied to vital loads considering single failures.
II. REVIEW CRITERIA The criteria for evaluating the loss of non-emergency ac power to the station auxiliaries are given in Standard Review Plan Section 15.2.6.
The objective is to verify that the plant responds to the most limiting transients in such a way that the criteria regarding fuel damage and system pressure are met.
III. EVALUATION Although a transient analysis of the loss of ac power was not performed as part of the orginal plant safety analyses, a discussion of loss of ac power and component failures in the auxiliary electric power system was provided in Section 9.1 of Reference 1. Two types of malfunctions were considered:
(1) malfunction in the switchyard, and (2) equipment malfunctions. In connection with the automation of the auxiliary feedwater system (AFWS), a transient analysis of the loss of non-emergency ac power to the station auxiliaries was performed.
The analysis results are provided in Appendix A of Reference 2. Features of the electric distribution system and the transient analyses are discussed in the following paragraphs.
A. Electric Distribution System Several types of malfunctions of the electric distribution system at San Onofre Unit 1 were discussed in Reference 1. These included malfunctions of the (1) switchyard, (2) turbine or reactor, (3) generator zone, (4) auxiliary transformer C, and (5) the 4,160 V buses.
Since the switchyard was modified since the description in Reference 1 (see References 3 and
- 4) these malfunctions are discussed in the following paragraphs.
- 1. Switchyard Malfunction Malfunction of any major item of equipment in the switchyard would be detected by the appropriate bus differential relaying system or breaker backup relay system. In either event, the faulty component would be electrically isolated. The switchyard is of the double breaker-double bus design and operates with all circuit breakers closed. A fault or failure in the switchyard, or tripping of the generator 220 kV circuit breakers will not result in a loss of power to the auxiliary electric system. Power would continue to be supplied without interruption through auxiliary transformer C.
-2 In the event of an internal fault in either of the two circuit breakers feeding auxiliary transformer C, both circuit breakers would trip and the power supply through transformer C would be lost. In this event, the 4,160 V buses will be aligned to receive offsite power through the main transformer and auxiliary transformers A and B by closing one or both of the circuit breakers from the control room-. The time required to restore power to the entire auxiliary electric power system from the initial malfunction is less than two minutes. If in addition there is a failure of the disconnect switch dc drive motor, this alignment will require manual opeation of the switch and increase the elapsed time for restoration of power to approximately seven minutes.
In the unlikely event that auxiliary power cannot be restored from the 220 kV system, one or both diesel generators may be placed in service.
- 2. Turbine or Reactor Malfunction Malfunctions in either the reactor or turbine cycle which trip the turbine will also open the generator circuit breakers. However, generator excitation is maintained to permit the turbine generator to supply coast-down energy to the reactor coolant pumps.
The pumps and exciter will operate for about five minutes with the speed being reduced to between 40 and 50 percent before their respective circuit breakers are tripped. Following coast-down, power is restored to buses 1A and 1B from the switchyard by closing tie breakers 11COl and 12C01 of the 4,160-V system.
Loss of generation from a reactor or turbine malfunction does not cause a loss of power in the switchyard. The offsite power system is always operated with sufficient spinning reserve to accommodate the transient which would result from an unexpected loss of power from the plant.
- 3.
Generator Zone Malfunction Malfunctions within the generator zone are detected by one or more relay systems; including differential, transformer sudden Pressure, overcurrent, stator ground, or loss of field.
Operation of any one of the relay systems would immediately trip the generator 220 kV circuit breakers and the 4,160 V circuit breakers, supplying buses 1A and lB.
Thus, immediately after such a malfunction, the turbine and reactor are tripped and the generator zone is electrically isolated.
Flywheel inertia of the reactor coolant pumps provides sifficient flow to prevent core damage.
Power would then be restored to buses 1A and 1B from auxiliary transformer C by closing tie breakers 11COl and 12C01.
At no time are buses 1C or 2C which supply the station auxiliaries without power.
-3
- 4.
Auxiliary Transformer C Trip A malfunction associated with auxiliary transformer C would be detected by one or more relay systems; including differential, highside overcurrent, and transformer sudden pressure.
The operation of any of these relay systems would immediately trip the 4,160 V circuit breakers resulting in loss of power to buses 1C and 2C and to the station auxiliaries.
In this event, power would be restored to these buses through the main transformer and auxiliary transformers A and B as discussed above for a switchyard malfunction.
This action will require less than two minutes, unless there is a concurrent failure of the disconnect switch dc drive motor, in which case seven minutes would be required. In addition, the diesel generators would be available.
- 5. Fault of 4,160 V Buses During operation, a bus fault on bus 1A or lB would be cleared by the overcurrent protection on circuit breakers 11A04 or 11B04.
During plant start-up, buses 1A and lB are fed from adjacent buses 1C and 2C, respectively, and the fault would be cleared by circuit breakers 11001 or 12001. A fault on bus 1A or 1B would not affect buses 1C or 2C except that during start-up, the bus feeding bus 1A or 1B would experience a momentary voltage dip which would last until the relays detect the fault and open tie breakers 11001 or 12C01.
Loss of power to bus 1A results in the loss of reactor coolant pumps A and C. Loss of power to bus 1B results in the loss of reactor coolant pump B and the generator exciter. Either or both of these incidents will result in turbine areactor trirping action except at reactor power levels less than 10 percent.
A bus fault on 4,160 V buses 1C or 2C would be cleared by the ovecurrent protection on circuit breakers 11C02 or 12CO2. Each of these buses feeds one complete safety injection train. Thus, loss of either bus still leaves the station with one complete safety injection train.
Loss of either bus IC or bus 2C would result in a loss of feedwater trip because of loss of one feedwater train.
B. Transient Analysis The initial stages of a loss of ac power transient resemble those of a loss of feedwater flow event. Later, after the reactor coolant pumps begin to coast down, the response is similar to a loss of forced coolant flow. Natural circulation is established and heat is removed by means of the steam generators.
The detailed analysis performed to simulate system transients is based on the LOFTRAN code developed by Westinghouse.
Three different cases were performed:
-4
- 1. Station blackout (loss of offsite power) with AFWS flow of 260 gpm at 3 minutes and 440 gpm at 8 minutes.
- 2. Station blackout with AFWS flow of 208 gpm at 8 minutes. Single failure of tie turbine-driven pump is assumed.
- 3. Station blackout with AFWS flow of 260 gpm at 3 minutes.
Single failure of the motor-driven pump is assumed. This case also models the loss of onsite and offsite ac power with no single failure, i.e., the turbine-driven pump and its associated flow path and instrumentation are independent of ac power.
Major assumptions and results of the analysis are provided below.
- 1. Analysis Assumptions The following assumptions are made in performing the analysis.
- a.
The plant is initially operating at 103 percent of rated power with three loops in operation.
- b.
Initial reactor coolant average temperature is 40F above the nominal value, and the initial pressurizer pressure is 30 psi above its nominal value.
- c. No credit is taken for the pressurizer power-operated relief valves or pressurizer spray.
- d.
Initial pressurizer level is at the nominal programmed value plus 2 percent (error); initial steam generator water level is at the nominal valve.
- e.
Main f:ewater to all steam generators is assumed t.o st;op at 10 seconds; i.e., 10 seconds of steady state operation is assumed.
- f. Reactor trip and auxiliary feedwater actuation are assumed to occur at 20 seconds (10 seconds into the transient).
Reactor trip from the steam/feed mismatch signal and auxiliary feedwater system actuation from the steam generator low water level signal would occur sooner based on current setpoints.
However, the delay is conservative in two respects. First, more energy is deposited into the reactor coolant prior to trip, which must be removed with the auxiliary feedwater system; second, the delay means a later auxiliary feedwater system start time.
- g. Auxiliary feedwater is assumed to be actuated at the flow rates and times listed above. The actuation delay times are referenced from the auxiliary feedwater actuation signal occurring 10 seconds into the transient. The system is assumed to be aligned such that each of the steam generators receives equal amounts of auxiliary feedwater.
00
-5 The 8 minute actuation delay time assumed for the motor-driven AFW pump takes into account the maximum time delay associated with operator action, according to current procedures, to attempt to restore offsite power and if unsuccessful to energize equipment from the onsite diesel generators, which is consistent with the design basis for the emergency onsite power supply.
The 3 minute actuation delay time assumed for the turbine-driven AFW pump allows time for operation of the turbine control valves, pump discharge control valves, and main steam drain and supply valves and time to preheat the new turbine rotor installed as part of the modifications for automatic initiation of the AFWS.
- h. A turbine trip is assumed to occur at the same time as the reactor trip.
- i.
No credit is taken for heat deposited in RCS metal during the RCS heatup.
- j.
No credit is taken for charging or letdown.
- k. Steam generator heat transfer area is assumed to decrease as the shell side liquid inventory decreases.
- 1. Conservative core residual heat generation is assumed based upon long-term operation at the initial power level preceding the trip.
- m.
An auxiliary feedwater temperature of 100OF is assumed.
- n.
A feedwater system purge volume of 73 ft3/1ooLis assumed.
This piping volume must be purged of the relatively hot main feedwater before the colder auxiliary feedwater enters the steam generators.
- o. Reactor coolant pumps are assumed to be tripped at the time of reactor trip.
- p. Secondary system steam relief is achieved through the steam generator safety valves.
- q.
No reactor control systems are assumed to function.
The reactor protection system is required to function following a station blackout as analyzed here.
No single active failure will prevent operation of this sytem.
-6
- 2. Analysis Results Results for the three station blackout cases are shown in Figures 1 to 15.
For all three cases, the pressurizer does not fill.
The auxiliary feedwater flows assumed are sufficient to remove core decay heat.
Since the reactor coolant pumps are tripped at the time of reactor trip, removal of pump heat is not necessary.
Prior to auxiliary feedwater actuation, each of the cases are identical.
Since the peak pressurizer water volume is obtained during this initial portion of the transient, each case has the same peak volume of 707 ft. 3 at 26 seconds.
IV. CONCLUSION The analysis assumptions for the loss of ac power are consistent with current criteria. In addition, in the analysis performed, operator action is assumed to occur at 8 minutes in the case of a single failure of the turbine driven pump. This operator action time is considered sufficient in view of the nature of the event (i.e., it would be straight forward to recognize based on alarms and indications) and the existing operating procedures.
The interruption of power does not jeopardize safe operations of the plant.
The secondary heat sink will remain effective in cooling the core.
The core will thus remain covered with water at all times, and therefore, in a coolable geometry to insure the fuel integrity. It is concluded that no core damage would occur as a result of such a transient and the analysis meets current criteria and is acceptable.
V. REFERENCES
- 1. San Onofre Unit 1 Final Safety Analysis Report, Section 9.1
- 2. Letter from K. P. Baskin to 0. M. Crutchfield dated March 6, 1981, titled "Basis for Auxiliary Feedwater System Flow Requirements, San Onofre Unit 1."
- 3. San Onofre Units 2 and 3 Final Safety Analysis Report, Section 8.2.
- 4. Letter from D. M. Crutchfield to R. Dietch dated February 6, 1981, Amendment No. 52 to Provisional Operating Licenses No. DPR 13.
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