ML20086C024
| ML20086C024 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 06/30/1995 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M92051, ULNRC-3234, NUDOCS 9507060240 | |
| Download: ML20086C024 (10) | |
Text
' d' 1901 Chouteau Avenue Post Office Box 149 St Louis. Missouri 63166 314-554 2650 i
e Donald F. SchneII i
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June 30, 1995 j
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 ULNRC-3234 Washington, D.C. 20555 TAC NO. M92051
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Gentlemen:
I DOCKET NUMBER 50-483 CALLAWAY PLANT OPTIMIEATION OF OVERPOWER DELTA-T REACTOR TRIP
Reference:
ULNRC-03198 dated April 17, 1995 In support of the referenced amendment application, additional information was transmitted by telephone facsimile on May 3 and May 9, 1995.
This additional information is formally transmitted as attached herewith.
If you have any questions on the attachment, please contact us.
Very truly y urs, Donald F. Schnell GGY/jdg i
b 9507060240 950630 PDR ADDCK 050004B3 g
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Donald F. Schnell, of lawful age, being first duly sworn upon oath says that he is Senior Vice President-Nuclear and an officer of' Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.
By
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D'onald F. Schnell Senior Vice President Nuclear SUBSCRI D and sworn to before me this O
day of 1995.
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5 MY COMMIS$10NEXPIRES APR.18,1998
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o cca T. A.
Baxter, Esq.-
Shaw, Pittman, Potts & Trowbridge 2300 N. Street, N.W.
Washington, D.C.
20037 M. H. Fletcher Professional Nuclear Consulting, Inc.
19041 Raines Dr Derwood, MD 20855-2432 M. J. Farber Chief, Reactor ProjectsSection III A U.S. Nuclear Regulatory Commission Region III 801 Warrenville Road Lisle, IL 60532-4351 Bruce Bartlett Callaway Resident Office U.S. Regulatory Commission RR#1 i
Steedman, MO 65077 i
t L. R. Wharton (2) l Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E21 11555 Rockville Pike Rockville, MD 20852-2738 i
Manager, Electric Department l
Missouri Public Service Commission P.O.
Box 360 Jefferson City, MO 65102 l
Ron Kucera Department of Natural Resources P.O. Box 176 I
Jefferson City, MO 65102 i
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u ULNRC-3234 L
Page 1 of 4 i
, Additional Information Sunnortina Chances to Callaway I
T/S 2.2.1 l
Background
There are essentially 3 changes being made to T/S 2.2.1:
is increased from In Table 2.2-1 Note 3, the value of K4 1.08 to 1.09.
The corresponding Allowable Value in Note 4 is reduced from 3.0% to 2.4% (calibrated span) and the Total Allowance is reduced from 5.7% to 5.0% (calibrated span).
Refer to Attachment 1 for an overview of the setpoint rack-up.
In Table 2.2-1 Note 3, the value of K6 is decreased from 0.0065 tof0.0015 (per. *F).
This parameter is used in the setpoint equation to compensate overpower protection for average temperatures greater than nominal Tavg.
The new value was determined by Westinghouse utilizing the OPTOAX code.
The revised setpoint equation will continue to ensure that core thermal and overpower limits are adequately protected.
See Attachment 2 (taken from WCAP-8746 for demonstration purposes only).
In Table 2.2-1 Note 1, a minor typographical error in the setpoint equation is corrected to obtain consistency with WCAP-7878, Revision 5.
The temperature compensation term on the right hand side of the equation was changed from
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'l + r S' l+rS K
(T-T']
to K
T - T' 2
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Since T' is a constant, the two expressions are mathematically equivalent.
Reasons For Changes During the previous cycle, as many as 7 alarms per day were experienced.
These alarms were an unnecessary operator distraction.
See Attachment 3.
Reactor power was limited to approximately.99.2% during the previous cy:le.
An increase to full power will increase the nuaber of alares and channel trips.
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ULNRC-3234 i
Page 2 of 4
* Duringlthe previous cycle, reactor power was' reduced-whenever OPAT channel 1, 3 or!4 was tripped for
. surveillance testing. _If a coincident trip occurred in
' channel 2~(which is most sensitive 1to.the upper head 7
anome.y' discussed below),:a reactor trip would.have 1
F resultaG.-
Reactor Flow Anomaly I
Callaway has had a flow anomaly in the upper plenum of the
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reactor vessel since Cycle-2.'
The anomaly is characterized by aperiodic, opposing step changes in loops 2.and 3 hot leg a
f temperatures.
The exact _cause of'the anomaly is unknown;
.however, it is apparently' associated with a flow
-distribution switching phenomenon in the reactor vessel' upper plenum. _During an anomaly transient, hotter water j
from the center of the core is redirected towards loop.2, j
while cooler water from the periphery of the core is redirected towards loop 3.
l.
Similar upper plenum flow anomalies (UFAs) have been experienced at other Westinghouse 4-loop PWRs.
At some i
-plants, theyanomaly is more severe, with hot leg temperature i
swings of 7'F occurring up to 80 times an hour as compared' l
to the 1.5 P swings occurring 3-4 times an hour at Callaway.
(Note that the frequency and magnitude of'the temperature swings vary from cycle to cycle-and are not necessarily constant during a particular cycle.)
The UFA'resulted in spurious.OPAT rod stop and' partial trip alarms during Cycle 7.
Other than these alarms, the anomaly i
did not cause any significant operational problems.
'j Furthermore, evaluations performed by Westinghouse and Union Electric have concluded that no nuclear safety or thermal cyclic fatigue concerns are associated.with the anomaly.
For a detailed description of the upper plenum flow anomaly, see Westinghouse letter ET-NRC-93-4015 dated November 12, i
1993 and WCAP-13778, " Analysis of Thermal Effects on Component Fatigue Resulting from the Upper Plenum Anomaly in Westinghouse PWRs".
M 2 OPAT M m-As' discussed above, the UFA resulted in spurious loop 2 OPAT rod stop and partial trip alarms at the start of Cycle 7.
During the period from 11/28/93 through 3/26/9,4, as many as 7 alarms / day were received.
The average alarm
-j rate during this period was approximately.1 alarm / day.
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ULNRC-3234 Page 3 of 4
'i
' Reanalysis of Full Power Steam Line Break Coincident with Rod Withdrawal While performing an evaluation to justify a change in the OPAT setpoint function, Westinghouse determined that it was necessary to reanalyze the full power steam line break coincident with rod withdrawal accident.
This reanalysis was prompted by the proposed change to Ks and a recent revision to the associated Westinghouse Safety Analysis Standard.
Detailed results of the updated analysis will be documented in a forthcoming FSAR changt, pending approval of the K6 change.
However, for the purpos'e of supporting the T/S 2.2.1 amendment, a summary of the shalysis is provided below.
Descriotion of Event The failure of main steam piping in the Turbine Building could produce adverse environmental conditions in the vicinity of the turbine impulse pressure transmitters.
.These transmitters provide control signals to the rod control system.
For the purpose of this analysis, it has been postulated that the effect of the steam line break is to cause an adverse failure of the pressure transmitters.
This failure would then cause a mismatch in Tavg and the controlling reference temperature, thus inducing a continuous control rod withdrawal.
Method of Analysis This transient was analyzed using the LOFTRAN computer code, which simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator and steam generator safety valves.
The code computes pertinent plant variables, including temperatures, pressures and power level.
A detailed thermal hydraulic computer code, THINC, was used to determine if DNB occurs for the core conditions computed by LOFTRAN.
Assumptions for the transient include the following:
a)
The accident was analyzed using the Improved Thermal Design Procedure.
Initial core power, RCS temperature, and RCS pressure were assumed to be at their nominal steady-state full power values.
Uncertainties in initial conditions were included in the DNBR limit, as described in WCAP-8567 (ITDP).
The RCS total flow rate was assumed to be the minimum measured flow.
s ULNRC-3234 Page 4 of 4 b)
The transient was analyzed with both minimum and maximum i
reactivity feedback.
2 c)
A range of break sizes was assumed (0.1 to l'.4 ft ),
d)
The highest worth RCCA was assumed to be stuck in its i
fully withdrawn position.
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Results of Most Limitina Case I
Upon initiation of the steamline break, control rods begin to withdraw.
This causes an increase in reactor power and core heat flux to the point at which the OPAT trip setpoint is reached.
Subsequently, a reactor trip occurs and control rods insert into the core.
The most limiting part of this transient occurs immediately prior to the reactor trip.
A DNB analysis was performed using the WRB-2 DNBR Correlation to determine that the DNBR was greater than the safety analysis limit values at all times during the transient.
Additionally, the peak linear heat rate and RCS pressure remained below their respactive limits.
Thus, all applicable acceptance criteria were satisfied and the analysis results were considered acceptable.
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