ML20238F095
| ML20238F095 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 08/24/1998 |
| From: | Barron H DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9809030108 | |
| Download: ML20238F095 (6) | |
Text
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h NM Duk3 Energy Corporation r# Energy.
uccuire Nuciear siation 12700 Hagers Ferry Road Huntersville, NC 28078-9340 i
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- 11. B. Barron (704) 875-4800 omCE l
Vice1%ident (704) 875-4809 Mx l
August 24,1998 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555
Subject:
McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Response to Questions Regarding Proposed Technical Specification (TS) Amendment TS 3.3.2 - Engineered Safety Features Actuation System Instrumentation By letter dated October 6,1997, McGuire Nuclear Station submitted the proposed TS Amendment to eliminate TS 3.3.2 Low Steam Line Pressure Safety Injection Function. By letter dated January 3,1997, Catawba Nuclear Station submitted a similar proposed TS Amendment.
By letter dated March 20,1997, Catawba responded to an RAI from the NRC dated March 17, 1997. By a telephone conference call on August 13,1998, the NRC requested McGuire to respond to the same questions included in the March 17,1997 RAI. Attached are the responses for McGuire.
This response contains no regulatory commitments.
Please contact P.T. Vu at (704) 875-4302 regarding this response.
Very truly yours, J
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8wk W
\\yd H.B. Barron r
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Attachment l
9809030108 980824 PDR ADOCK 05000369-P png
I US NRC-August 24,1998 Page 2 xc:(w/ attachment)
L.A. Reyes Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta St., NW, Suite 2900 Atlanta,GA. 30323 F. Rinaldi Senior Project Manager Office of U.S. Nuclear Reactor Regulation One White Flint North, Mail Stop 14E21 Washington, D.C. 20555 S.M. Shaeffer NRC Senior Resident Inspector L
- McGuire Nuclear Station R.M. Fry, Director Division of Radiation Protection State of North Carolina 3825 Barrett Drive Raleigh, N.C. 27609-7221 l
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bxc: (w. attachment)
ELL (EC050)
M. Kitlan (CN0lRC)
L.B. Jones (EC050)
P.T. Vu (MG0lRC) 4 T. Niggel (EC08H)
M. Weiner (MG0 LOP)
K. Crane (MG01RC) i i
US NRC-Attachment August 24,1998 Page 1 of 3 QUESTIONS AND RESPONSE REGARDING j
ELIMINATION OF LOW STEAM LINE PRESSURE SAFETY INJECTION
- 1. Does the removal of the low steam pressure safety injection (SI) signal impact the emergency operating procedures? If so, please explain.
Response
Yes, Emergency Procedures associated with a reactor trip or safety injection have a step to check and determine if safety injection has occurred automatically. If not, a series of questions is asked to see if automatic injection should have occurred. One of the parameters checked is " main steam pressure above 775 lbs?" With the steam pressure SI signal deleted, this criterion will be removed.
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- 2. It is the staff's understanding that by removing the low steam pressure safety injection signal, you intend to prevent an SI actuation following a loss of offsite power event.
How have you ensured that the resultant change in the SI actuation could effectively suppress an unnecessary SI actuation in this case?
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Response
l No, the purpose of the proposed change is to eliminate spurious safety injections. The proposed change was one of several options analyzed, and is considered to be the most likely to successfully reduce the likelihood of future unnecessary actuations, with the attendant challenges to equipment, and thermal transients.
- 3. Discuss your plan to take corrective actions on the design deficiency in the main steam system to prevent the potential excess reactor coolant system cooldown following a loss of offsite power event. The proposed plant configuration could lead to a transient different from the event analyzed in Final Safety Analysis Report.
Response
A plant modification was implemented to facilitate failing open the miscellaneous main steam line drain valves on loss of instrument air, however maintaining them closed following a single unit LOOP provided instrument air is still available. A modification is currently in progress to add diesel powered instrument air compressors. Diesel powered instrument air l
will keep the drain valves closed and also allow for more timely control of auxiliary feedwater flow during a dual unit LOOP.
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1 US NRC' Attachment August 24,1998 Page 2 of 3 Additionally, revisions were made to the reactor trip emergency procedure. These include:
- 1) Guidance was added earlier in the emergency procedure to assure that if 6.9 KV power l
was not available (indicating a LOOP), auxiliary feedwater (CA) flow should be i
throttled as appropriate, based on steam generator (SG) level. (Earlier guidance to close the main steam isolation valves (MSIVs) and the MSIV bypass valves was added l
here previously.)
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- 2) Generic steps for throttling CA to stop an uncontrolled cooldown, which appear in I
several places in the EPs, were clarified to better express the option of throttling CA flow regardless of SG level, as long as CA flow is maintained above the CA minimum flow value.
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- 4. What spectrum of break sizes was evaluated for the main steamline break transient with the removal oflow steam pressure safety injection signal? Please provide detailed results of your evaluation.
Response
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2 For the BWI feedring steam generators, break sizes ranging from 0.5 ft to 4.5 ft were 2
l analyzed in increments as small as 0.1 ft in order to ensure a conservative result. A listing of break sizes and their effect on peak core heat flux is presented in the response to Question #6.
- 5. You stated in your submittal that a small steamline break causes the SI to actuate on low pressurizer pressure before reaching the setpoint of the low steam pressure safety injection signal. At what break size does the Si actuation switch from low pressurizer pressure to low steam pressure?
Response
For the BWI feedring steam generators, the SI actuation switches from low pressurizer 2
pressure to low steam line pressure for break sizes greater than 2.5 ft,
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from low steam pressure. Include the sequence of events and the major transient l
curves (nuclear power, pressure, temperature, DNHR, etc.).
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Response
l The limiting break size is that which produces the highest peak core heat flux. Previous 2
analyses with SI on low steam line pressure showed that the limiting break size is 2.0 ft for i
US NRC*
Attachment August 24,1998 Page 3 of 3 the BWI feedring steam generators. The results shown below for the offsite power maintained case demonstrate that this same break size still produces the hignest peak heat Dux with the removal of SI on low steam line pressure. Since the limiting break size results in low pressurizer pressure SI actuation prior to reaching the low steam line pressure SI setpoint, the peak heat flux is not affected by the removal of SI on low steam line pressure for the limiting case. For the larger non-limiting break sizes, SI actuation on low pressurizer pressure results in an increase ofless than 2% in peak heat flux when compared to the transient response with SI actuation on low steam line pressure.
BWI feedring steam generators:
Peak core heat Oux (%FP)
Break size without SI on low steam line pressure 2
0.5 ft 19.09 2
1.5 ft 27.80 2
1.9 ft 29.28 2
2.0 ft 29.45 2
2.1 ft 39,44 2
2.5 ft 29.35 2-3.5 ft 28.92 2
4.5 ft 28.45 Note that because the limiting break size is unaffected by the deletion of the low steamline pressure signal, all of the sequence of events and transient curves presented in Chapter 15 of the station UFSAR remain valid.
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