ML20138A497

Forwards Response to 970411 Telcon for Clarification of Details of License Amend Request for Steam Generator Tube Rupture Evaluation

ML20138A497
Person / Time
Site:	Catawba
Issue date:	04/16/1997
From:	Mccollum W DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M98107, TAC-M98108, NUDOCS 9704280090
Download: ML20138A497 (4)

Text

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, DukeIbuerCompany WruntR hkoutht, JR

'Cotau ba Nuclear Generatnn D<partment

, Vi e haident 4800 Concordfand (M3)MIJ200 Office nrk. SCBin (M3)UIJLYifar

( DUKEPOWER April 16, 1997 U. S. Nuclear Regulatory Commission j Attention: Document Control Desk j Washington, D.C. 20555 I

j

Subject:

Catawba Nuclear Station, Units 1 and 2

) Dockets Nos. 50-413 and 50-414 l Request for Additional Information Regarding the

! Operating License Amendment for the Steam Generator Tube Rupture Evaluation (TAC Nos. M98107, M98108)

Per phone conference on April 11, 1997, the NRC requested clarification of details regarding the additional information submitted by Duke Power to the.NRC in letters dated April 2, 1997 and April 10, 1997 for the Steam Generator PORV Technical Specification Amendment. The proposed amendment revises section 3/4.7.1.6 of the

. Technical Specifications to require four instead of three steam generator power operated relief valves (PORVs) and Section 15.6.3 of the Updated Final Safety Analysis Report (UFSAR) to require four instead of three PORVs and allow credit for local manual operation of the PORVs. The additional information requested is provided in the enclosure and should supply the clarification necessary to complete the amendment request.

We request that you review the additional information on a schedule consistent with the urgency of the original request. If you need additional clarification of the response to the questions or have additional questions please contact Martha Purser at (803)-831-4015.

Sincerely, William R. McCollum, Jr. , p 9704280090 970416 IT , . , N' INS.

PDR ADOCK 05000413;  %

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.i Document Control Desk

, Page 2 l April 16,1997 1

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L.A. Reyes, Regional Administrator, Region II 4

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P.S. Tam, Senior Project Manager, ONRR l

R.J. Freudenberger, Senior Resident Inspector, CNS 1

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Clarification of Request for Additional Information for Catawba Uniti and Unit 2 Regarding Proposed Amendments for the Steam Generator PORVs

Reference:

Phone Conference with the state NRR branch on April 11, 1997 and April 14, 1997 regarding letters from Duke Power to NRC dated April 2, 1997 and April 10, 1997 Question 1:

Why are the initial condition uncertainties (10% and 13.6%)

associated with steam generator (SG) narrow range level different than the uncertainty (4 %) associated with throttling of auxiliary feedwater on level indication?

Response

At a full power condition, SG levels are automatically controlled to a program level setpoint based on indicated reactor power. Thus, the uncertainty associated with indicated reactor power is a componant of the overall uncertainty associated with the initfal SG narrow range level uncertainty. Another component factored into the overall initial SG narrow range level uncertainty is the velocity head effect. This effect is caused by high liquid / steam flow passing over the pressure taps used to measure SG level. The overall SG narrow range level uncertainty at a full power condition is 10% for Unit 1 and 13.6% for unit two.

Following reactor trip, SG Tevel is manually controlled.

Thus, the uncertainty associated with indicated reactor power is not factored into the overall SG narrow range uncertainty. The liquid / steam flow passing over the pressure taps used to measure SG level is very low during post-trip conditions, such that the velocity head effect uncertainty is negligible. When these components are factored out of the overall SG narrow range level uncertainty for post-trip conditions, the resultant level uncertainty is 4%.

Question 2:

In the Unit 2 overfill analysis, does reactor trip occur on low RCS pressure (1945 psig) or on safety injection actuation (1845 psig) ? Was uncertainty applied to this setpoint?

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Response

Reactor trip occurs on low RCS pressure in the Unit 2 overfill analysis. Positive uncertainty was applied to this setpoint such that the length of time for reactor trip to occur is minimized.

Question 3:

The table in response to question 8a in the April 2, 1997 letter to the NRC does not have assumed or observed times under the action for Throttling of CA. A time of 10 minutes is assumed in the WCAP-10698. What are the times?

Assumed times are noted in the Summary Tables for overfill Analysis in the letter to the NRC dated April 10, 1997. Unit 1 is 5.9 minutes, Unit 2 is 11.5 minutes. The actual time on the simulator is 11.5 minutes. This difference is due to the assumptions regarding the initial levels in the steam generator. The analysis assumes the maximum '.evel plus positive level uncertainty. The simulator starts with a realistic operating level which allows a longer time to throttle CA.

Question 4:

This question requests additional information on time trials and the percentage of operators trained.

Response

The sole purpose of running the time trials was to validate

.the times for the Chapter 15 analysis. Approximately three trials were run. The times in the analysis are the documented, verifiable information. The Steam Generator Tube Rupture SGTR Accident is one of the simulator events on which the operators are regularly tested. The operations procedures have not changed therefore no additional training is-required. The actual manual operation of the PORVs has been reviewed by every operator in the last several weeks.

This information is discussed in the April 2, 1997 letter from Duke Power to the NRC.

The use of sampling to verify SGTR is still not used at Catawba. This is noted in the April 2, 1997 letter (question 9, #5).