ML20137G599

Forwards Responses to Staff Questions Re Proposed TS to Eliminate Safety Injection Signal on Low Steam Pressure

ML20137G599
Person / Time
Site:	Catawba
Issue date:	03/20/1997
From:	Tuckman M DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M97538, TAC-M97539, NUDOCS 9704010441
Download: ML20137G599 (6)

Text

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-11 I u Dukelbterr Company MShm l

P.0 Box 1006 ,

Senior Vice President

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, , Garlotte,NC282011006 NuclearGeneranon (704)392200 Othce (704)3824360 Fax DUKEPOWER March 20, 1997 U. S. Nuclear Regulatory Commission Washington, D. C. 20555  :

Attention: Document Control Desk l,

Subject:

Catawba Nuclear Station Docket Numbers 50-413 and -414  !

Proposed Technical Specification to Eliminate Safety (

Injection Signal on Low Steam Pressure; Response to RAI (TAC Nos. M97538 and M97539) ,

By letter dated January 3, 1997, Duke Power Company requested a license' amendment that would remove actuation of safety injection i on receipt of a low steam line pressure signal. By letter dated  ;

March 17, 1997, the NRC requested additional information.

Attached are responses to the Staff's questions.  ;

7 If any additional information is required, please call Scott Gewehr at (704) 382-7581.

bd-M. S. Tuckman cc: Mr. P. S. Tam, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O-14 H25 Washington, D. C. 20555 Mr. L. A. Reyes, Regional Administrator i

U. S. Nuclear Regulatory Commission - Region II  !

101 Marietta Street, NW - Suite 2900 Atlanta, Georgia 30323 l l

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1 9704010441'970320 PDR ADOCK 05000413 p PDR

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I U .* S . Nuclear Regulatory Commission ,

March 20, 1997- 1 Page 2 R. J. Freudenberger Senior Resident Inspector Catawba Nuclear Station 6

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Responses to Questions on Deletion of SI on Low SteamLine Pressure )

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1. Does the removal of the low steam pressure safety injection (SI) signalimpact the emergency operating procedures? If so, please explain.

Response

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 j series of questions is asked to see if automatic injection i 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.

Other procedure changes, to-ameliorate design deficiencies identified in the February 6, 1996 loss-of-offsite power /SI event, are discussed in response to Question #3. .

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2. It is the staffs 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

The opportunity for an unnecessary SI signal is not limited to a LOOP event. (Watts Bar recently suffered an unexpected actuation while cooling down in Mode 4.) The proposed modification 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.

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I Response ,

Several changes or clarifications were added to the emergency procedure for a reactor trip as a result of the February 6, 1997 LOOP event, as well as a December 27, 1993 event at McGuire. These include:

1) Former Step 4, check for 6.9 KV Power, was moved up in priority to become Step 2. (The two steps

, bypassed were administrative; announcement of reactor trip, and determination of classification i

level.)

2) Guidance was added to this step to assure that if 6.9 KV power was not available (indicating a LOOP),

I auxiliary feedwater (CA) should be throttled as appropriate, based on steam generator (S/G) level.

(As a result of the December 1993 McGuire event, i a step to close the main steam isolation valves (MSIVs) and the MSIV bypass valves was added here.)

3) The generic steps for throttling CA, which appear in many places in the EPs, were clarified to better

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express the option of throttling CA regardless of  !

S/G level, as long as CA flow is maintained above '

the CA minimum flow value.

4. What spectmm 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

For the Model D steam generators (Catawba Unit 2, and McGuire Unit 2 until September, 1997), break sizes ranging from 0.4 ft2 to 5.4 ft2 were analyzed in 1 ft2 increments (per the approved methodology of DPC-NE-3001). For the BWI feedring steam generators, break sizes ranging from 0.5 ft 2 to 4.5 ft2 were analyzed in increments as small as 0.1 ft2 in order to produce a more 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 Model D steam generators, break sizes greater than 1.4 ft2 result in a low steam line pressure SI prior to reaching the setpoint for SI on low pressurizer pressure.

For the BWI feedring steam generators, the SI actuation switches from low pressurizer pressure to low steam line

-pressure for break sizes greater than 2.5 ft 2,

6. Provide the results of the reanalyzed steamline break transient without SI initiated from low steam pressure. Include the sequence of events and the major transient curves (nuclear power, pressure, temperature, DNBR, etc.).

Response

The limiting break size is that which produces the highest peak core heat flux. Previous _ analyses with SI on low steam line pressure showed that the limiting break sizes are 1.4 ft2 for the Model D steam generators and 2.0 ft2 for the BWI feadring steam generators. The results shown below for the offsite power maintained case demonstrate that these same break sizes still produce the highest peak heat fluxes 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 I 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 of.less than 2%

in peak heat flux when compared to the transient response '

with SI actuation on low steam line pressure.

Model D steam generators:

Peak core heat flux (%FP)

Break size without SI on low steam line pressure 0.4 ft" 16.76 1.4 ft 2 25.89 2.4 ft 2 25.66 3.4 ft 2 24.71 4.4 ft 2 24.28 5.4 ft 2 23.38

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BWI feedring steam generators:

Peak core heat flux (BFP)

Break size without SI on low steam line pressure 0.5 ft" 19.09 1.5 ft 2 27.80 l 1.9 ft 2 29.28 2.0 ft 29.45 2.1 ft 2 29.44 ,

2.5 ft 2 29.35 I 3.5 ft 2 28.92 )

4.5 ft 2 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 stations' UFSARs remain valid. I i

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