ML20214W184
| ML20214W184 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 11/24/1986 |
| From: | Ainger K COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 2423K, NUDOCS 8612100118 | |
| Download: ML20214W184 (3) | |
Text
f3
/
\\ Commonwealth FAIson
_ }
/\\ ) One First Nabonel Plaza. Chcago, IEnois m.
Address Reply to: Post Omco Box 757 U Chicago. IEnois 60600 - 0767 November 24, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC. 20555
Subject:
Byron Station Units 1 and 2 Application for Amendment to l
Facility Operating License NPF-37, Appendix A. Technical Specifications NRC Docket Nos. 50-454 and 50-455
References:
(a) "05000454/LER-1985-061, :on 850624,reactor Trip Occurred Due to Steam Generator 1A LO-2 Level.Caused by Attendant Inadvertent Activation of Local Overspeed Trip Bar for [[system" contains a listed "[" character as part of the property label and has therefore been classified as invalid. Pump 1C.Trip Lever Handles Removed|November 26, 1985 letter]] from K.A. Ainger to H.R. Denton (b) June 16, 1986 letter from K.A. Ainger to H.R. Denton
Dear Mr. Denton:
References (a) and (b) requested a license amendment which would revise Technical Specification 3.5.2, Emergency Core Cooling Systems. The proposed amendment would allow certain valves in the safety injection system to be temporarily closed during leak testing of some reactor coolant system pressure isolation check valves.
Attachment A of this letter contains additional information to support the proposed technical specification change.
One signed original and fifteen (15) copies of this letter and attachment are provided for NRC review.
Please direct any questions regarding this matter to this office.
Very truly yours, K.A. Ainger Nuclear Licensing Administrator
/klj att.
cc: Byron Resident Inspector Y
4
\\ \\
86 21 % $ h
-e ATTACHMENT A The amendment being requested would allow either valve SI8809A or B to be closed concurrently with the SI8835 valve to perform leakage testing of certain reactor coolant system (RCS) pressure isolation check valves per Specification 4.4.6.2.2.
During discussions with the NRC staff regarding this proposed amendment, a question was raised regarding the ability of the emergency core cooling system (ECCS) to respond to a loss of coolant accident (LOCA) while the check valve leakage surveillance was being performed.
During this surveillance, the unit would be in Mode 3 at a reduced RCS pressure of approximately 1000 psi with one SI8809 valve and the SI8835 valve closed simultaneously for up to two hours. This configuration would limit the available ECCS flow to mitigate the effects of a LOCA.
A large spectrum of LOCA break sizes has been evaluated in the FSAR accident analysis for Mode 1 conditions. However, Mode 3 conditions at approximately 1,000 psi are far below the conditions for which the reactor coolant system has been designed that a large LOCA is not credible and for all practical purposes can be assumed not to occur.
Engineering studies, leak-before-break analysis and operating experience have shown through wall cracks in the RCS Class 1 pressure boundary piping, greater than six inches in diameter, are highly unlikely. Therefore, for purposes of this discussion, a credible LOCA of piping less than six inches was evaluated.
A shutdown LOCA scoping study was performed by Westinghouse for a four-loop plant two hours af ter a reactor trip (reference Westinghouse June 24, 1986 letter NS-NRC-86-3144).
Based on this study, it was estimated that at least 20 minutes would be available to initiate safety injection (SI) flow from a centrifugal charging pump to prevent significant core uncovery for breaks up to three inches in diameter. Initiation of SI flow within this time may not preclude core uncovery, but is expected to limit the fuel cladding heatup to less than the full power small break LOCA results provided in the FSAR. For breaks larger than three inches and up to six inches in diameter, operator action to initiate SI from a centrifugal charging pump is estimated to be required within approximately 10 minutes. Additional operator action may be required, depending upon break size, within one hour of the event initiation to start an additional centrifugal charging pump or safety injection pump or depressurize the RCS using the steam generators and to start a RHR pump.
Several factors should be considered regarding the performance of this-check valve leakage surveillance which would significantly increase the time required for operator action from the times identified in the scoping study discussed above.
i l
-__ -. _ _ -. -. -. = - _
I J.
e Several indications are available to the operator to identify that a LOCA is in progress. These include:
loss of pressurizer level, RCS pressure 4
decrease, loss of RCS subcooling, radiation alarms inside the containment, containment pressure increase and sump water level increase. When a LOCA has been identified..the operator would initiate a safety injection as required by the emergency procedures. Because of the plant configuration during the check valve leakage surveillance, when a SI is initiated, ECCS flow would automatically be available from two centrifugal charging pumps, one RHR pump injecting into two cold legs and the safety injection accumulators.
Since this surveillance would be performed in Mode 3, the RHR pumps suction would be aligned to the RWST. The scoping study assumed flow from only one centrifugal charging pump, so it is evident that a substantial amount of additional ECCS flow would be available.
In addition, the emergency response procedures direct the operator to verify ECCS flow. Since these valves could only be closed for up to a two hour period, the operator would be aware these valves are closed and would open them to restore all available ECCS flow.
Simple manual action by the operator in the control room can reopen valves SI8809 and SI8835.
4 i
The scoping study assumed the LOCA occurs two hours after a reactor trip. This check valve surveillance will be performed during startup 1
following a shutdown. Typically, this would be after a refueling outage or after the plant has been in Mode 5 for longer than three days, much longer than the two hours assumed in the scoping study. Therefore, the surveillance would be conducted when the initial fuel rod temperature and decay heat levels are less than those assumed in the scoping study. This would also extend the times before operator action would be required to longer than the times i
resulting from the scoping study.
. Normal operating pressure in Mode 1 serves as a more severe condition which demonstrates that pipe ruptures below normal operating pressure are highly unlikely since additional margins of safety exist at the lower pressure. The condition that could lead to a pipe rupture, a large through wall crack, would be identified during operation. However, even with the presence of such a crack, the piping system would remain stable and a pipe rupture would be unlikely at the reduced RCS pressure.
i In summary, a LOCA occurring during Mode 3 conditions at reduced RCS pressure is highly unlikely. A LOCA occurring in Mode 3 during the short period of time SI8809 and SI8835 valve are closed concurrently is even more unlikely. However, if a LOCA did occur, the preceding discussion demonstrates j
that sufficient time would exist for operator response given the plant configuration.
2423K i
4 I
i
- - - -. - - - - - - - - -