ML20207D710

Forwards Request for Addl Info Re Snupps Steam Generator Tube Rupture Analysis.Info Requested within 45 Days of Ltr Receipt,To Permit Continued Review of 860211 & 0401 Submittals on Current Schedule

ML20207D710
Person / Time
Site:	Wolf Creek, Callaway, 05000000
Issue date:	07/15/1986
From:	Oconnor P Office of Nuclear Reactor Regulation
To:	Koester G KANSAS GAS & ELECTRIC CO.
Shared Package
ML20207D676	List: ML20207D671 ML20207D710
References
NUDOCS 8607220166
Download: ML20207D710 (4)

Text

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i . . " "' a g y, ~.' g UNITED STATES 3 '*^ n NUCLEAR REGULATORY COMMISSION 5 ' , *; j WASHINGTON, D. C. 20555 g . .v ..+

",/ 15 JUL 1986 i

Docket No.: 50-483 Mr. Glen L. Koester Vice President - Nuclear Kansas Gas & Electric Company 201 North Market Street Wichita, 63166

Subject:

Request for Additional Infonnation Related to the SNUPPS Steam Generator Tube Rupture Analysis The staff is continuing its review of your submittals dated February 11, 1986 and April 1,1986. To pennit us to continue our review on our current schedule, we require the infonnation requested in_ Enclosure I to this letter be provided.

Please provide the requested infonnation within 45 days of your receipt of this letter.

Sincerly, 0- W Paul W. O'Connor, Project Manager PWR Project Directorate #4 Division of PWR Licensing-A

Enclosure:

As stated I

cc: See next page 8607220166 860715 2 PDR ADOCK 0500 P

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<s Mr. Glenn L. Koester Wolf Creek Generating Station i Aansas Gas and Electric Company Unit No. I cc:

Mr. Nicholas A. Petrick Mr. Gary L. Haden. Director Executive Director, SNUPPS Research & Energy Analysis 5 Choke Cherry Road Kansas Corporation Commission Rockville, Maryland 20850 4th Floor - State Office Building Topeka, Kansas 66612-1571 Jay Silberg, Esq.

Shaw, Pittman, Potts & Trowbridge Regional Administrator, Region IV 1800 M Street, NW U.S. Nuclear Regulatory Commission Washington, D.C. 20036 Office of Executive Director for Operations Mr. Donald T. McPhee 611 Ryan Plaza Drive, Suite 1000 Vice President - Production Arlington, Texas 76011 Kansas City Power & Light Company 1330 Baltimore Avenue Mr. Allan Mee Kansas City, Missouri 64141 Project Coordinator Kansas Electric Power Cooperative,Inc.

Chris R. Rogers, P.E. P. O. Box 4877 Manager, Electric Department Gage Center Station Public Service Commission Topeka, Kansas 66604 P. O. Box 360 Jefferson City, Missouri 65102 Resident Inspector / Wolf Creek NPS c/o U.S. Nuclear Regulatory Commission Regional Administrator, Region III P. O. Box 311 U.S. Nuclear Regulatory Commission Burlington, Kansas 66893 799 Roosevelt Road Glen Ellyn, Illinois 60137 Mr. Brian Moline Chief Legal Counsel Brian P. Cassidy, Regional Counsel Kansas Corporation Commission Federal Emergency Management Agency 4th Floor - State Office Building Region I Topeka, Kansas 66612-1571 J. W. McConnack POCH Boston, Massachusetts 02109 Senior Resident Inspector / Wolf Creek NPS c/o U.S. Nuclear Regulatory Commission 3 Mr. Robert Elliot, Chief Engineer P. O. Box 311 Utilities Division Burlington, Kansas 66839 i- Kansas Corporation Commission 4th Floor - State Office Building Topeka, Kansas 66612-1571 Mr. Gerald Allen Public Health Physicist Bureau of Air Quality & Radiation Control Division of Environment

Kansas Department of Health l and Environment Forbes Field Building 321 Topeka, Kansas 66620 l

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Enclosure (1)

Request for Additional Information Steam Generator Tube Rupture (SGTR) Analysis for SNUPPS Plants

1. Please justify the conservatism of the break flow model utilized. In Appendix D it is stated that 3 correlations may be applicable: (a)

Modified Zaloudek; (b) Burnell; (c) Henry, 1970 Model. It was concluded that the Burnell correlation most conservatively estimated the critical flow rates for the range of conditions encountered'in the SGTR event and this correlation was presumably utilized for the break flow

• rate, providing an initial break flow rate of 44 lb/sec (Figures 4-8 and 4-23). However, in Appendix C (Verification of RETRAN and Scoping Code),

the RETRAN break flow model was modified to use the modified Zaloudek correlation, since this correlation is used in the FSAR, and in accordance with Figure C-8 an initial break flow rate of 70 lb/sec was obtained.

Please clarify this apparent discrepancy.

Note: A 50% increase in break flow rate, as suggested in Figure C-0, would be more than enough to flood the main steam lines to the ARV location and would probably completely fill the main steam system (see also Figure

' . 4-11). Since the margin to overfill for Case 1 is only 91 cubic feet, even a small increase in break flow would result in overfill.

2. Justify the use of the scoping code for sensitivity studies since a comparison of SGTR results utilizing the scoping code versus RETRAN shows

' significant differences for break flow rates, primary and secondary pressures (See Figures C-1, C-3 and C-4).

3 ., The RETRAN Model for the. pressurizer utilizes a single-node

, non-equilibrium fluid volume with phase seperation. Potential shortcomings of such a model are: (1) the non-equilibrium model tends to over-predict the pressurizer pressure; (2) the single node representation assumes full mixing of the hot resident fluid with the cooler insurging primary coolant during the SI phase, and thus ignores stratification, causing underprediction of the pressure, It is not clear how these effects interact. The RCS pressure in the SNUPPS analysis appears to stay j considerably higher than in other SGTR analyses for Westinghouse plants available to the staff. Therefore, please justify the pressurizer model by comparison with multi-node models and experimental data.

4. The secondary side of the steam generators is modeled as a single saturated volume. The staff is particularly concerned about the adequacy of the model for the case of the " stuck open ARV", which results in very low steam generator levels (see Figure 4-26). Therefore, please justify the model adequacy by comparison with: (1) results of analyses using multinode steam generator secondary models which have the capability to predict stratification; (2) experimental data.

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5. As noted above, the " stuck open ARV" case shows very low steam generator levels (equivalent to a mixture volume of about 1500 cu. ft.). For this case you assumed that a Hot-leg tube rupture located at the top of the tube sheet would result in the greatest offsite dose. The staff believes that the worst case would be a rupture of the topmost tube at the U bend, since the U bend would probably be above the mixture level for at least part of the transient, and would have the greatest potential for iodine transport to the steam soace even when immersed. Therefore, please provide adequate justification that the break location chosen for this case is the most conservative from an offsite dose standpoint, or provide the results of calculations assuming the break occurs at the U bend.

6. Please provide the basis for the assumption that primary and secondary pressure equalization is achieved within 5 minutes after SI termination.

This is not apparent from a comparison of primary and secondary pressures (Figures 4-1and4-9). How is pressure equalization accomplished?

7. Please explain whether the RETRAN model includes the pump loop seals.

This is not apparent in Figure B-1. Also provide the elevation differences utilized in the model.

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