ML20062D897
| ML20062D897 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 08/02/1982 |
| From: | Fiedler P GENERAL PUBLIC UTILITIES CORP. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-05.B, TASK-03-06, TASK-03-10.A, TASK-05-05, TASK-06-07.A3, TASK-06-07.C1, TASK-06-10.A, TASK-07-01.A, TASK-08-02, TASK-08-04, TASK-3-10.A, TASK-3-5.B, TASK-3-6, TASK-5-5, TASK-6-10.A, TASK-6-7.A3, TASK-6-7.C1, TASK-7-1.A, TASK-8-2, TASK-8-4, TASK-RR NUDOCS 8208060257 | |
| Download: ML20062D897 (7) | |
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s GPU Nuclear Corporation NMC MF 100 Interpace Parkway Parsippany. New Jersey 07054 201 263-6500 TELEX 136-482 Writer's Direct Dial Number:
August 2, 1982 Mr. Dennis M. Crutchfield Chief, Operating Reactors-Branch No. 5 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Sir:
SUBJECT:
Oyster Creek Nuclear Generating Station Docket No.-50-289 5, 6 - 0 J 7 System Evaluation Program During the period of May 18 through May 20, 1982, GPUN committed to providing additional information in many SEP topic areas.
Listed below, by topic, are some of the items you have been expecting.
III - 5.B It has been verified that both the turbine building closed cooling water and fire protection systems operate at less than 150 psig and 200*F.
In consideration of these parameters, it can readily be seen that a rupture of these sytems is extremely unlikely.
III - 6 As-built drawings for the affected portion of the CRD piping have been completed and given to the NRC staff.
III - 10 Th identification of the 12 valve motor operators with bypassed thermal overloads is per Enclosure 1.
The identification of all remaining ESF valve motor operators whose power is supplied from ESF motor control centers will be provided by September 1982.
The evaluation planned to provide resolution of this topic is described in Enclosure 2.
This evaluation will be completed by December 1983.
V-5 The evaluation to make containment radioactive gas and particulate monitors operational will be completed by December 1982.
Completed tables of Reg. Guide 1.45 are attached.
VI - 7. A.3 Surveillance procedures for the ECCS actuation systems have been provided to the NRC staf f.
8208060257 820802 PDR ADOCK 05000219 p
PDR GPU Nucleai Corporation is.i subsidiary of the General Pubhc Utthties Cuporation
c VI - 7.C.1 The study to verify proper selective coordination of protective devices to and from the AC system ABT buses to demonstrate that no single failuro of a protective device would disable redundant safety features will be completed by December 1982.
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VI - 10.A The review of existing plant surveillance procedures to ascertain whether safety functions affected by the reactor mode switch are tested will be completed by December 1982.
VII - 1.A The failure modes and effects analysis to evaluate the need for isolation d evices be tween the nuclear flux monitoring systems and their recorders will be completed by December 1982.
VIII - 2 It has been verified that an undervoltage protective trip does not exist for the diesel generators.
VIII - 4 The evaluation concluding that th re exists adequate assuranc e against seal failure on the typical medium voltage penetration due to a fault current is described in Enclosure 3 The evaluation of the adequacy of the backup pro tective device clearing times to provide protection against seal failure on the typical low voltage penetration due to fault current will be completed by December 1983 The additional information that you require conc erning topics III-6, VI-4, VI-7. A.3, VII-1A and VIII-2 will be provided as it becomes available.
In the event that any comments or questions arise, please contact Mr. J. Knubel of my staff at (201) 299-2264.
Very truly yours, Peter B. Fiedler Vice President and Director Oyster Creek Enclosure s j
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N Topic III-10.A Thermal OL Protection for Motor Operated Valves The 12 valve motor operators with bypased thermal overload relays referred to in the NRC contractors report #1653F dated April 1980 are:
1)
V-20-3 Core Spray Pump Suction Valve - System 1 2)
V-20-4 Core Spray Pump Suction Valve - System II
- 3) V-20-12 Core Spray Pump Discharge Valve - System I
- 4) V-20-15 Core Spray Isolation Valve - System I
- 5) V-20-18 Core Spray Pump Discharge Valve - System II
- 6) V-20-21 Core Spray Isolation Valve - System II
- 7) V-20-26 Core Spray Recire. Valve - System II
- 8) V-20-27 Core Spray Recire. Valve - System I
- 9) V-20-32 Core Spray Pump Suction Valve - System I
- 10) V-20-33 Core Spray Pump Suction Valve - System II
- 11) V-20-40 Core Spray Isolation Valve -System I
- 12) V-20-41 Core Spray Isolation Valve - System II References Burns & Roe drawing 3020
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f Topic III-10.A Thermal OL Protection for Motor Operated Valves Evaluation for Establishing OL Setpoints An evaluation shall be conducted on the application of thermal overload relays on a case-by-case basis for each ESF valve motor operator to determine if setpoints can be established with all uncertainties resolved in favor of completing the safety related action and still provide meaningful protection.
Consideration shall be given to:
a) variations in ambient temperatures b) inaccuracies in the overload device c) set point drift d) the effect on motor current of lower than nominal voltage due to degraded grid conditions.
e) the cumulative effect of heating caused by any possible successive starts at short intervals.
If thermal overload relay setpoints cannot be so established, the thermal overload may be bypassed continually (except for test) or bypassed only during accident conditions.
The evaluation and resulting recommendations shall ensure that the pro-tective device coordination is such that these devices do not unjustifiably compromise the completion of the safety function or degrade other safety systems because of sustained locked rotor or fault currents that may exist.
-...A Docket No. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION SEP TOPIC VIII - 4 ELECTRICAL PENETRATIONS OF REACTOR CONTAINMENTS May 1982
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Docket No. 50-219 SEP TOPIC VIII - 4 Comments in response to NRC Letter dated March 26, 1981 and NRC Contractor's Report dated January 15, 1980.
The January 29, 1981 letter by Northeast Utilities Company delineated their concerns over the model employed in the NRC contractor's evaluation, and we concur in that expression of concern.
The following represents clarification of and additional information to that provided to the NRC on this topic in the JCP6L letter (Finfrock) dated April 24, 1979.
TYPlCAL MEDIUM VOLTAGE PENETRATION - JCP&L has previously identified penetration #20 (CE NS03) as being typical of medium voltage penetrations.
This penetration pro-vides 4160 volt AC power to Reactor Recirculation Pump Motor NC01-B.
This pene-tration uses 500 MCM cable with a continuous current rating of 475 amperes.
The primary protective device is an instantaneous overcurrent relay, GE type 12PJC31F23A connected in a differential mode. The secondary protective device is an instantaneous overcurrent relay, GE type 12PJCllAV3A.
In differential protection a vectorial comparison is made of the currents on each side of the zone of protection in each phase.
Any current dif ference (dif ferential current) above the relay setting causes the relay to produce a trip.
Normal load current or current due to a fault external to the zone of protection produces equal currents at each side of the protection zone and no differential current flows to the relay.
Current due to a fault within the zone of protection produces currents vectorially unequal at each side of the protection zone resulting in a differential current flow to the relay and producing a trip.
For the typical medium voltage penetration referenced, the differential zone of protection in-cludes the generator windings of the MG set, the cable to the penetration, the penetration circuit, the cable from the penetration to the Reactor Recirculation Pump Motor, and the windings of the Reactor Recirculation Pump Motor.
Maximum short circuit current at the penetration has previously been determined by JCP&L to be 1800 amperes.
Both primary and secondary protective devices would operate for a fault at the penetration and trip the feeder circuit breaker to the motor of the MG set and trip the field circuit breaker of the generator of the MG set.
At maximum short circuit current, the NRC contractor's evaluation, using its highly conservative model, calculated that penetration seal over-temperature would be reached in 98 seconds from initial LOCA temperature.
How-ever, fault current at the penetration is not produced from the offsite electri-cal distribution network but by the generator of the Reactor Recirculation Pump MG set, and the maximum value of fault current persists only for several seconds as the generator effective impedance changes from its transient value to its synchronous value.
The fault current present after several seconds is well below the continuous current rating for this penetration. Therefore, even if both the primary and secondary protective devices fail to clear a fault at the penetration, the penetration seals are inherently protected from overtemperature due to fault current magnitude because of the characteristic change in generator impedance.
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This demonstrates that there exists adequate assurance against seal failure on the typical medium voltage penetration due to fault current.
TYPICAL LOW VOLTAGE PENETRATION - JCP&L has previously identified penetration #11 (GE type NSO4) as being typical of low voltage penetrations. This penetration provides 460 volt AC power to the Drywell Recirculation Fan RF-1-1.
The pene-tration uses #2AWG cable and has a continuous current rating of 66 amperes.
Maximum short circuit current has previously been determined by JCP&L to be 4200 amperes. At maximum short circuit current, the NRC contractor's evalua-tion, using its highly conservative model, calculated that the primary protective device would operate prior to seal failure but that the secondary protective de-vice clearing time would be too long.
Review of the adequacy of backup protective device clearing times to provide protection against seal failure on the typical low voltage penetration due to fault current is still under investigation.
Should it be determined to be in-adequate for penetration seal protection per the NRC position, possible correc-tive action would be either replacement of certain circuit breakers on feeders to MCC buses with units having different tripping characteristics or installation of fuses or current limiters in certain power circuits.
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