5N 1578 Lookout Place

NOV 03 1989

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

Gentlemen:

In the Matter of Tennessee Valley Authority

) Docket Nos. 50-327 ) 50-328

SEQUOYAH NUCLEAR PLANT (SON) AND WATTS BAR NUCLEAR PLANT (WBN) LINE BREAKS (MSLBs) IN ICE CONDENSER PLANTS

References:

MAIN STEAM

1. NRC letter to TVA dated March 3, 1989, "Main Steam Line Breaks in Ice Condenser Plants - Sequoyah, Units 1 and 2 (TAC R00199, R00200) and Watts Bar Units 1 and 2"

2. TVA letter to NRC dated August 17, 1989, "Sequoyah Nuclear Plant (SON) and Watts Bar Nuclear Plant (WBN) - Main Steam Line Breaks in Ice Condenser Plants"

Reference 1 transmitted to TVA the staff contractor's draft report summarizing the COMMIX computer code results for an MSLB event in an ice condenser containment. The staff solicited TVA's comments on the draft report, particularly on the section in which the contractor noted that certain local containment temperatures (as calculated by CrIMMIX) exceeded environmental qualification temperatures for SQN.

TVA's comments were provided in Reference 2. In general, TVA contends that the COMMIX results are overly conservative when applied to local containment temperatures. This is the result of the modeling abilities of the COMMIX code. The COMMIX results do, however, support the overall containment response as calculated by LOTIC III.

During an August 24, 1989, telephone conference call, Jack Donohew and Chang Li of NRC requested that additional information be provided on the 4SLB issue regarding the relationship between SQN and WBN plant configuration and the COMMIX and COB6-.-NC analyses treatment of the break location. This additional information is provided as an enclosure.

891114(-A04 891103 PDR AT' CK 05000327 S'!C

An Equal Opportunity Employer

50-390 50-391

TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 37401

U.S. Nuclear Regulatory Commission NOV 03 1989

No commitments are contained in this submittal. Please direct qt'estions concerning this Issue to Russell R. Thompson at (615) 843-7470.

Very truly yours,

TENNESSEE VALLEY AUTHORITY

Manag),Nuclear Licensing and Regulato--y Affairs

Enclosure cc (Enclosure):

Ms. S. C. Black, Assistant Director for Projects

TVA Projects Division U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852

Mr. B. A. Wilson, Assistant Director for Inspection Programs

TVA Projects Division U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323

NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379

NRC Resident Inspector Watts Bar NJclear Plant P.O. Box 700 Spring City, Tennessee 37381

* ENCLOSURE

Main Steam Line Breaks (MSLBs) in Ice Condenser Plants Sequoyah Nuclear Plant (SQN), Units 1 and 2 and Watts Bar Nuclear Plant (WBN), Units 1 and 2

The major difference between the COBRA-NC analysis performed by Westinghouse Electric Corporation and the Argonne National Laboratory COMMIX confirmatory analysis was the temperature in the Immediate vicinity of the break location. One of the factors contributing to the difference in the break node temperature identified by the Argonne confirmatory analysis was the difference in the treatment of break location by COMt4IX and COBRA-NC. The staff requested TVA to examine the plant configuration to determine which model more accurately represented the expected response of the containment to an MSLB.

In order to perform this evaluation, pipe failure studies that addressed General Design Criterion 4 requirements of Appendix A of 10 CFR 50 were reviewed for SQN and WBN. The postulated failure locations predicted by the pipe break studies (References I and 2) are located inside the vertical guard pipes surrounding the main steam line and atop the steam generator (see Figures 1 and 2). These break locations are consistent among all steam generators in both plants.

An examination of SQN and WBN drawings of the main steam line and an informal walkdown of WBN were performed. The segment of the main steam line enclosed in the vertical guard pipe, for which breaks are postulated, is located approximately 1/4 distance between the crane wall and the biological shield wall (see Figure 3). The postulated breaks inside the guard pipe and atop the steam generator will yield downward-facing flow into the lower compartment. Because of the main steam line guard pipe and the steam generator doghouse geometry, flow behavior in the vicinity of these postulated breaks will result In the entrainment and mixing of surrounding colder fluid away from the crane wall in the lower compartment. This is consistent with the modeling in the COBRA-NC analysis.

The formulation of the governing equations In COBRA-NC and COMMIX involving mass and energy transfer was examined for differences that would affect mixing In the region of the break. The COBRA-NC formulation of both the transverse and vertical momentum equations contains terms that address Interfacial drag between the vapor and liquid phases and momentum exchange because of mass transfer between phases (Reference 3). The COMMIX formulation uses a one mixture momentum equation (Reference 4) that does not contain these terms. The presence of these additional terms in COBRA-NC will result In different mixing profiles than predicted by the COMMIX formulation. In addition, use of a mixture formulation instead of a multiphase formulation can potentially result In vaporization of drops in the presence of the high jet temperature. The result is that instead of entraining flow from all directions as is typical for a jet, nonphysical flow would be out of the break cell on all sides of the jet. Therefore, the overall cell temperature in the region of the break would nonphysically approach the jet temperature. A comparison of the COBRA-NC results (Figures 6.14 and 6.23 of Rcference 5) and COMMIX results (Figure 4.57 of Reference 6) illustrates this phenomena.

In conclusion, higher temperatures in the vicinity of the break calculated by the COMMIX model are the result of physical mixing phenomena inconsistent with expected break location and droplet entrainment. The COMMIX models were intended to be only bounding calculations with the results to be used for verification purposes only. The results provided by the more rigorous COBRA-NC model more accurately represent mixing phenomena and temperatures near postulated MSLB locations.

References:

1. Sequoyah Nuclear Plant, Units 1 and 2, "Protection Against Dynamic Effects of Pipe Failure Inside Containment and the Main Steam Valve Rooms," Tennessee Valley Authority, Report No. CEB-76-3

2. Evaluation c1 the Effects of Postulated Pipe Failures Inside Containment and the Main Steam Valve Rooms for Watts Bar Nuclear Plant, Units 1 and 2, Report No. CEB-79-39, Tennessee Valley Authority

3. NUREG/CR-3262, "COBRA-NC: A Thermal-Hydraulic Code for Transient Analysis of Nuclear Reactor Components," Volume 1, Pacific Northwest Laboratories, May 1986

4. NUREG/CR-4348, "COMMIX-lB: A Three-Dimensional Transient Single-Phase Computer Program for Thermal Hydraulic Analysis of Single and Multicomponent Systems," Volume 1, Argonne National Laboratory, September 1985

5. Westinghouse Electric Corporation, WCAP-10988, "COBRA-NC, Analysis for a Main Steamline Break in the Catawba Unit 1 Ice Condenser Containment." Hochreiter, L. E., El. Al, November 1985

6. Letter from Suzanne Black (NRC) to 0. D. Kingsley, Jr., (TVA) dated March 3, 1989, "Main Steam Line Breaks in Ice Condenser Plants - Sequoyah, Units 1 and 2 (TAC R00199, R00200) a;id Watts Bar Units I and 2"

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