ML20040G866
| ML20040G866 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway, 05000000 |
| Issue date: | 02/04/1982 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SLNRC-82-07, SLNRC-82-7, NUDOCS 8202160554 | |
| Download: ML20040G866 (9) | |
Text
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SNUPrS Standardized Nuclear Unit Power Plant System 5 Choke Cherry Road Nicholas A. Petrick Rockville, Meryiend 20850 Executive Director (301) 8694010 February 4,1982 SLNRC 82 07 FILE: 0541 SUBJ:
NRC Request for Additional Information - Reactor Systems Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C.
20555 Docket Nos: STN 50-482 and STN 50-483
References:
1.
NRC'(Youngblood) letter to KGE (Koester), dated December 31, 1981:
Same subject 2.
NRC (Youngblood) letter to UE (Schnell), dated January 15, 1982:
Same subject
Dear Mr. Denton:
The referenced letters requested additional information on the SNUPPS plants concerning the analyses.of a locked reactor coolant pump rotor and a sheared pump shaft. The enclosure to this letter provides the requested information and will be incorporated in Revision 8 to the SNUPPS FSAR.
Very truly yours, b
Nicholas A. Petrick i
g RLS/mtk4b28&4bl6-18 g$ECENED 9
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Enclosure t
FEB1'81982> ~E cc:
G. L. Koester KGE f afearusumensam '
D. T. McPhee KCPL e,'
smuinawata
//
D. F. Schnell UE T. E. Vandel NRC/WC
'g J. H. Neisler NRC/ CAL 6%
8202160554 820204 PDR ADOCK 05000482 A
SNUPPS 440.1 The analyses of a locked reactor coolant pump rotor and a sheared (440.3WC)
(440.lC) reactor coolant pump shaft in the FSAR assumes the availability of offsite power throughout the event.
In accordance with Standard Review Plan 15.3.3 and GDC 17, ve require that this event be analyzed assuming turbine trip and coincident loss of offsite power to the undamaged pumps.
Appropriate delay times may be assumed for loss of offsite power if suitably justified.
Steam generator tube leakage should be assumed at the rates specified in the Technical Specifications.
The event should also be analyzed assuming the worst single failure of a safety-system active component. Maximum technical specification primary system activity and steam generator tube leakage should be assumed. The analyses should demonstrate that offsite doses are less than 10 CFR 100 guidelines values.
RESPONSE
Accident Scenario The locked Rotor followed by a loss of offsite power transient is postulated to occur in the following manner:
Reactor coolant pump rotor locks (or_ shears) and flow in that loop 'begins to a.
coastdown.
b.
The reactor is tripped on low RCS flow in one loop.
c.
Turbine / Generator trips.. -..
d.
Offsite power is lost.
NOTE: Grid stability analyses show that the grid will remain stable and offsite power will not be lost because of a unit trip from 100% power. Refer to Section 8.2.2 of each Site Addendum. The following analysis assumes a 2 second time delay between reactor trip and loss of offsite power. This is a conservative assumption based on the grid stability analyses.
e.
The loss of offsite power causes the three remaining reactor coolant pumps to coast down.
Method of, Analysis The method of analysis used is the same as the cases presented in Section 15.3.3.
The following case is analysed; Four loops operating, one rotor locks.
Followed by coastdown of other three reactor coolant pumps.
Results Figures 440.1-1 through 440.1-5 show a comparison between the locked rotor transient without offiste power and the locked rotor transient with offiste power from section 15.3.3.
As can be seen from the figures, losing offsite power results in the same peak clad temperature and the same peak RCS pressure.
The calculated sequence of events for the case without offsite power is shown in Table 440.1-1.
Conclusion The locked rotor without offsite power transient is no more limiting than the case preserted in Section 15.3.3..
-TABLE 440.1-1 4
. SEQUENCE OF EVENTS LOCKED ROTOR WITHOUT OFFSITELPOWER
. EVENT
-TIME-(SECONOS)
Rotor on one. pump locks 0.0 Low RCS flow trip setpoint reached
.05 Rods begin to drop 1.05 Maximum RCS pressure occurs 3.0 Maximum clad temperature occurs 3.01 Remaining reactor coolant pumps begin to coastdown 3.05' i-i I
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