ML20028G416
| ML20028G416 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/28/1983 |
| From: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8302090319 | |
| Download: ML20028G416 (6) | |
Text
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BALTI M ORE GAS AND ELECTRIC CHARLES CENTER.P.O. BOX 1475 BALTIMORE, MARYLAND 21203 ARTHUR E. LUNDVALL. JR.
vlCE PRESIDENT January 28,1983 Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. R. A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Subject:
Calvert Cliffs Nuclear Power Plant Unit 2, Docket No. 50-318 Testing of Third Train Auxiliary Feedwater System Reference (A): A. E. Lundvall to R. A. Clark letter dated 11-17-82,
" Supplement 1 to Fifth Cycle License Application" Gentlemen:
During testing of the new Auxiliary Feedwatcc (AFW) System, the failure of an AFW control valve to function properly promrced a review and discus::lon with NRC staff. NRC staff posed several questione related to the supplement to Cycle 5 l, reload license application, Reference (A). Responses to those questions are l contained in the attachment to this letter.
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- Should you have any questions, please contact us.
l l Very truly y e .
A. E. LuncVall Vice President, Supply AEL:fld Atta ment ec: J. A. Biddison, Esquire G. F. Trowbridge, Esquire D. H. Jaffe - NRC R. R. Mills - CE R. E. Architzel- NRC/CC h*\
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Office of Nuclear Rnactor-Rsgulation January 28, 1983
- Page 2 ATTACHMENT
- 1. Question Provide a discussion of the failure of the auxiliary feedwater control valve to operate properly.
Response
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A' discussion of this incident will be included in the followup report to LER 83-03/1T due January 26,1983.
- 2. Question Does the incident have any safety significance?
Response
An analysis confirmed our judgement that this potential for higher than normal AFW flow simultaneous with the most limiting overcooling design basis event would yield acceptable results. Twelve hundred (1200) gpm is the maximum possible AFW flow. AFW flow is limited to this value by the size of AFW pumps suction lines.
The Steam Line Break (SLB) event represents the worst case postulated overcooling event which could lead 'o a return to power. In thi:: SLB analysis, the effects of increased AFW flow upon the peak reactivity and peak return-to-power were compared with those of Reference (A1).-. AFW _ flow was assumed to continue at a rate of 1200 GPM for more than ten minutes after initiation. The analysis also assumed loss of AC power upon turbine trip from full power and the worst case stuck CEA (i.e., that CEA which yields the most severe combination of scram worth and reactivity-insertion is stuck fully l withdrawn). As with Reference (A1), credit was taken for the negative reactivity effects considered in HERMITE methodology. The peak post-trip reactivity is calculated to be -0.032% delta Rho, which is only 0.012% delta Rho less negative than the similar analysis of Reference (A1). The peak return to power is 8.97% (of 2700 MWT), which is only 0.37% greater than Reference (A1). Therefore, the conclusions of Reference (A1) remain valid, even with the higher AFW flow.
The analysis assumptions and initial conditions are identical to those listed in Reference (A1). A detailed description of the AFW system modeling is given below. Those aspects of the modeling which differ from those used in Reference (A1) are highlighted by an asterisk (*).
An auxiliary feedwater actuation analysis setpoint of 60.9% of steam generator wide range span is assumed in this analysis. This represents a Technical Specification actuation setpoint of 45.0% and includes a 15.9%
uncertainty. The actuation signal activates a motor driven auxiliary feedwater pump and a steam driven auxiliary feedwater pump which deliver auxiliary feedwater to both steam generators. For the case of loss of AC power en turbine trip, the motor driven pump auxiliary feedwater flow reaches the steam generator 43.8 seconds after the low steam generator level signalie .aated. This is the minimum time delay associated with the motor i
Office of Nuclear Racetor Rsgulation January 28, 1983 Page 3. -
driven pump to accelerate to' full speed (4.5 seconds) and other signal
, processing delay time (5.0 seconds),10.0 seconds for the diesel generator to start and reach speed following loss of AC, and 15.0 seconds (30.0 seconds if -
LOCA sequencer actuated due to a SIAS) for the motor driven auxiliary feedwater pump -to be loaded on line. The steam driven pump's auxiliary feedwater reaches full flow to the steam generator 9.5 seconds 'after the auxiliary feedwater actuation setpoint is reached. This includes a minimum time delay of 5.0 seconds required to open the steam admission valve to the auxiliary feedwater pump and 4.5 seconds for the pump to fully accelerate to speed once the admission valves are fully open. The steam driven pump is accelerating over the entire 9.5 second period. The analysis conservatively assumes that the auxiliary feedwater flow legs are filled with water and thus no time delay associated with auxiliary feedwater flow through the piping was included in the analysis.
During the acceleration of each pump, flow increases linearly from 0 gpm to 1200 gpm over the acceleration time (4.5 seconds for motor driven pump,9.5 seconds for steam driven pump). The flow control valves are assumed to fall in the full open position *. Therefore, once any AFW. pump is up to speed, system flow is assumed to be 1200 gpm and remains at 1200 gpm for a minimum of 10 minutes *.
The analysis also includes isolation of the ruptured steam generator when the steam generator differential pressure reaches the analysis setpoint of 250.0 psid. This represents a Technical Specification setpoint of 135.0 psid and an uncertainty of 115.0 psid. In addition, a 20.0 second time delay is assumed in the analysis to close the AFW isolation (i.e., block) valves. These assumptions are conservative since they delay the isolation of AFW to the ruptured steam generator.
It was further assumed that until automatic isolation occurs, all AFW flow is directed to the ruptured steam generator. No credit is taken for any throttling effect by the block valves.
Results The sequence of events is given in Table 2-1. This sequence is almost identical to that given in Reference (A1). The transient minimum DNBR was calculated to be 1.30 at 140.0 seconds.
- Conclusions The results of the Steam Line Break event with a failure of the AFW flow control valve to full open are consistent with and nearly identical to those in Reference (A1). Therefore, the conclusions in Reference (A1) are still valid.
- 3. Question What is the basis for the 160 gpm setpoint for the AFW Flow Control Valves?
Response
An equipment setpoint of 160 + 10 gpm for each AFW flow control valve ensures that there 'll not be excessive undercooling or overcooling of the
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Office of Nucle r Razetor R:gulation January 28, 1983 Page.4 reactor upon initiation of auxiliary feedwater. Specifically, a flow setpoint of at least 150 gpm ensures -that there is sufficient auxiliary feedwater to preclude loss of heat sink even without operator action for 10 minutes during the most limiting undercooling event, Feedline Break. See Reference A1. A flow setpoint of no more than 170 gpm also ensures that auxiliary feedwater flow for ten minutes will not result in actuation of safety injection should AFW flow be actuated as a result of the loss of one (1) main feed pump during 100% power operation. As discussed in the response to Question 2 above, the:
limiting overcooling event is the Steam Line Break (SLB) Accident. The SLB yields acceptable results for total AFW flow of 1200 gpm (1012 gpm including instrument loop uncertainties); or an equivalent of 300 gpm (253 gpm including instrument loop uncertainties) per flow control valve.
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l Reference (A1): A. E. Lundvall to R. A. Clark letter dated 11/17/82,
" Supplement 1 to Fifth Cycle License Application" TABLE 2-1 l
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Offics of Nuclear Rsactor Rsguletion January 28, 1983 Page 5 SEQUENCE OF EVENTS FOR INSIDE CONTAINMENT STEAM LINE BREAK EVENT WITH LOSS OF AC POWER ON TURBINE TRIP INITIATED FROM HFP .
(AND A FAILED OPEN AFW FLOW CONTROL VALVE)
Time (Sec.) Event ,,
Getpoint or Value 0.0 Steam Line Break Occurs 6.305 ft2 .
1.7 Low Steam Generator Pressure Analysis 600.0 psia Trip Setpoint is Reached; Steam Generator Isolation Analysis Setpoint is Reached 2.6 Trip Breakers Open; Main Steam Isolation Valves Begin to Close; Main Feedwater Valves Begin to Close 2.9 Steam Generator Differential Pressure delta P = 250.0 psid Analysis Setpoint is Reached 3.1 CEAs Enter Core; Loss of AC Power on
- Turbine Trip; RCPs Coastdown Begins; Diesel Generator Start Coming on Line; Main Feedwater Rampdown Begins
.13.1 Diesel Generators Reach Kated Speed Following LOAC Power; Shutdown Sequencer Initiated 13.4 Auxiliary Feedwater Actuation Analysis 60.9%**
Setpoint is Reached; Steam Driven Pump Begins to Accelerate as Steam Admission Valves Begin to Open 14.6 Main Steam Isolation Valves Completely Close 17.7 SafetyInjection Actuation Analysis 1645.0 psia Setpoint is Reached; LOCA Sequencer Initiated 18.4 Steam Admission Valves to Steam Driven AFW Pump Completely Opein 19.7 Pressurizer Empties i
- % of distance between steam generator v;ide range upper and lower level instrument taps.
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9 Office of Nuclear Rimetor Rcgulation )
. January,28,-1983
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~1 Time (Sec.) Event Setpoint or Value I
22.7 Power Provided to High Pressure Safety Injection Pumps
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22.9 AFW Block Valve Completely Closed; 1200 gpm Steam Driven AFW Pump at Full Speed; AFW Flow Directed to Intact Steam Generator 23.1 Main Feedwater Rampdown Completed 8% of Full Power Feedwater Flow 52.7 High Pressure Safety Injection Pump at Full Speed 57.2 Motor Driven AFW Pump at Full Speed 82.6 Main Feedwater Isolation Valve Completely Closed 100.5 Affected Steam Generator Blows Dry 132.5 Peak Reactivity -0.032% Delta Rho 133.0 Peak Return to Power 8.97% of 2700 MWt i
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