ML20112G512

Discusses Proposed Changes to post-fuel-loading Initial Test Program.Proposed Change to FSAR Chapter 14 Re Measured Flow Coastdown Does Not Involve Unreviewed Safety Question & Remains Valid

ML20112G512
Person / Time
Site:	Catawba
Issue date:	01/11/1985
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8501160364
Download: ML20112G512 (4)

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DUKE POWER GoxvAxy e.o. nox casso OHAHLOTTE, N.C. 28242 HALB. TUCKER TF.uPHOME vara p.mamewe (70-3) 373-4303 m... .-,,o January 11, 1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Waahington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: Catawba Nuclear Station Docket Nos. 50-413 and 50-414

Dear Mr. Denton:

By letters dated December 17, 1984 and January 2, 1985, Duke Power Company requested that the NRC Staff review and approve certain proposed changes to the post-fuel-loading initial test program for Catawba Unit 1. On January 3, 1985, representatives from Duke Power, Westinghouse and the NRC Staff discussed the safety analysis for the Reactor Coolant System flow coastdown which had been submitted by the January 2, 1985 letter. As a result of this discussion, Duke Power was requested to provide additional discussion to demonstrate that the measured RCS flow coastdown would not have an adverse impact on other FSAR Chapter 15 events. The following discussion is provided to supplement the above referenced letters.

In the FSAR analysis, 5 events are analyzed assuming a flow coastdown as calculated by the LOFTRAN (Reference 1) computer code. These are:

15.1.5 Steam System Piping Failure 15.2.6 Loss of Non-Emergency AC Power to the .ccation Auxiliaries 15.2.8 Feedwater System Pipe Break 15.3.3 Reactor Coolant Pump Shaft Seizure (Locked Rotor) 15.6.3 Steam Generator Tube Rupture The effect of the as measured flow coastdown on each transient is discussed below.

15.1.5 Steam System Piping Failure The analysis presented in FSAR Section 15.1.5 is performed to demonstrate that

"...the core remains in place and intact. Radiation doses do not exceed the guidelines of 10 CFR 100."

"Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable, the following analysis, in fact, shows that no DNB occurs for any rupture..."

Reference'It. Burnett, T.W.T., et al., "LOFTRAN Code Description," WCAP-7907,

. June 1972.

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Mr.-'Hard1d R.'Denton, Dircctor January 11, 1985 Page Two-At the limiting time in the DNB transient for the case without offsite power available the RCS flow is in natural circulation. Natural circulation flow is insensitive to the actual flow coastdown used to achieve natural circulation

, conditions. Thus the FSAR is not affected by the measured flow coastdown.

15.2.6 Loss of Non-Emergency AC Power to the Station Auxiliaries This event'is analyzed "to show that the natural circulation flow in the RCS following a Im , of a-c power event is sufficient to remove residual heat from the core." Although this is an ANS condition II event, DNB is not the primary concern. This is because in a true loss of a-c power analysis (i.e.

reactor coolant pump trip at initiation of event), the first few seconds of the transient would be similar to the complete loss of flow incident. The effect of the measured flow coastdown data on the complete loss of flow analysis has been previously reported to be acceptable with respect to meeting the DNBR limits.

During the time when natural circulation is of interest (3300 seconds after initiation of event), the RCS flow is independent of the time it takes for the flow to be in the natural circulation range. Thus the conclusion stated in the FSAR remains valid, i.e., sufficient natural circulation flow is available to provide adequate core decay heat removal following trip and RCP coastdown.

15.2.8 Feedwater System Pipe Break The feedwater system pipe break is performed to demonstrate that the assumed Auxiliary Feedwater System capacity'is adequate to remove decay heat, to prevent overpressurizing the RCS, and to prevent uncovering the reactor core.

FSAR Figure 15 2.8-6'shows that a steam bubble is present in the pressuriser x throughout the transient-(that is'the pressurizer does not go water solid) therefore the core remains covered. The as measured flow coastdown would not alter the presence of~a steam bubble in the pressurizer.

For the case without'offiste' power.available, the FSAR states ,

"...there will be a flow coastdown until flow in the loops reaches the natural circulation.value. The natural circulation capability of.the RCS-has been shown'in Section 15.2.6... to be~sufficlint to remove' core decay heat'following reactor trip."

During the time when natural circulation is of interest (4300 seconds after initiation of event), the RC8' flow is independent of the time it takes for the flow to be in the natural circulation range. Thus'the conclusion stated in the FSAR remains valid, i.e.,' sufficient natural circulation flow is available to g' provide adequate core decay heat removal following trip and RCP coastdown.

,15.3.3 Reactor Coolant Pump Shaft Seisure (Locked Rotor)

The locked rotor event is' analyzed without offsite power as. reported in Request

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Mr. Harald R. Denton, Dir:cter

January 11,.1985 Page Three for Additional Information 440.128. The response states that "the locked rotor without offsite power transient is no more limiting that the case presented in Section 15.3.3" (with offsite power).- This conclusion remains valid even when the measured flow coastdown data is considered.

For the Catawba FSAR analysis, no rods were predicted to experience DNB during '

a locked rotor event with or without offsite power available. This remains true when the measured flow coastdown is taken into account because the time of minimum DNBR occurs approximately 1 second prior to the time that RCP coastdown begins. Thus the measured flow coastdown data has no impact on the number of rodo predicted to experience DNB.

15.6.3 Steam Generator Tube Rupture The results of the Catawba flow coastdown test do not have any effect on the Catawba Steam Generator Tube Rupture (SGTR) analysis DNBR result as presented in Chapter 15.6.3 of the FSAR. For the analysis, loss of offsite power and RCP trip is assumed to occur coincident with reactor trip. The LOFTRAN results for Catawba show that DNBR increases after trip due to the reduction in power coincident with reactor trip having a greater effect than the flow reduction  ;

due to RCP coastdown. Therefore a slight reduction in coastdowt flow will not affect minimum DNBR for the SGTR accident analysis.

The DNBR for the SGTR analysis is bounded by the results for Section 15.6.1,

'" Inadvertent Opening of a Pressurizer Safety or Relief Valve." This is due to the more severe RCS depressurization from this transient. In Section 15.6.1 it was determined that DNBR'is always greater than 1.3 and thus no clad damage is expected. Thus, no clad damage is expected for the steam generator tube rupture accident.

Conclusion

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As noted in each of the above analyses, the measured flow coastdown does not adversely affect the current FSAR analysis. Therefore, our previous conclusion, ,

that the proposed change to FSAR Chapcer 14 does not involve an unreviewed safety question, remains valid.

Very truly yours, O cd k Hal B. Tucker ROS sib

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Mr. Har:1d R. Denton, Direct:r January 11, 1985 Page Four cc: Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Catawba Nuclear Station Robert Guild, Esq.

P. O. Box 12097 Charltston, South Carolina 29412 Palmetto Alliance 2135h Devine Street Columbia, South Carolina 29205 Mr. Jesse L. Riley-Carolina Environmental Study Group 854 Henley Place Charlotte,' North Carolina 28207 Ik '