ML20207D671
| ML20207D671 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway, 05000000 |
| Issue date: | 07/15/1986 |
| From: | Oconnor P Office of Nuclear Reactor Regulation |
| To: | Schnell D UNION ELECTRIC CO. |
| Shared Package | |
| ML20207D676 | List: |
| References | |
| NUDOCS 8607220158 | |
| Download: ML20207D671 (5) | |
Text
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o Docket No.: 50-482 Mr. Donald F. Schnell 15 JUL 1986 Vice President - Nuclear Union Electric Company Post Office Box 149 St. Louis, Missouri 63166
Dear Mr. Schnell:
Subject:
Request for Additional Infomation 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 permit us to continue our review on our current schedule, we require the information requested in Enclosure 1 to this letter be provided.
Please provide the requested information within 45 days of your receipt of this letter.
- Sincerly, 16\\
Paul W. O'Connor, Project Manager PWR Project Directorate #4 Division of PWR Licensing-A
Enclosure:
As stated cc: See next page
, DISTRIBUTION:
- NOTE:
Duplicate letter and enclosure
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also sent to: Mr. G. L. Koester NRC PDR '
Kansas Gas & Electric Co.
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15 JUL 1986 Docket No.: 50-482 Mr. Donald F. Schnell Vice President - Nuclear Union Electric Company Post Office Box 149 St. Louis, Missouri 63166
Dear Mr. Schnell:
Subject:
Request for Additional Infomation 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 pemit us to continue our review on our current schedule; we require the infomation requested, in Enclosure 1 to this letter be provided.
Please provide the requested information within 45 days of your receipt of this letter.
Sincerely, u
N Paul W. O'Connor, Project Manager PWR Project Directorate #4 Division of PWR Licensing-A
Enclosure:
As stated cc: See next page
O Mr. D. F. Schnell Callaway Plant Union Electric Company Unit No. I
+
cc:
Mr. Nicholas A. Petrick Lewis C. Green, Esq.
Executive Director - SNUPPS Green, Hennings & Henry 5 Choke Cherry Road Attorney for Joint Intervenors Rockville, Maryland 20850 314 N. Broadway, Suite 1830 St. Louis, Missouri 65251 Gerald Charnoff, Esq.
Thomas A. Baxter, Esq.
Ms. Marjorie Reilly Shaw, Pittman, Potts & Trowbridge Energy Chaiman of the League of 1800 M Street, N. W.
Women Voters of Univ. City, M0 Washington, D. C.
20036 7065 Pershing Avenue University City, Missouri 63130 Mr. J. E. Birk Assistant to the General Counsel Mr. Donald Bollinger, Member Union Electric Company Missourians for Safe Energy Post Office Box 149 6267 Delmar Boulevard St. Louis, Missouri 63166 University City, Missouri 63130 U. S. Nuclear Regulatory Comission Mr. Dan I. Bolef, President Resident Inspectors Office KayTrey, Representative RR#1 Board of Directors Coalition Steedman, Missouri 65077 for the Environment St. Louis Region Mr. Donald W. Capone, Manager 6267 Delmar Boulevard Nuclear Engineering University City, Missouri 63130 Union Electric Company Post Office Box 149 St. Louis, Missouri 63166 Chris R. Rogers, P.E.
Manager - Electric Department 301 W. High Post Office Box 360 Jefferson City, Missouri 65102 Regional Administrator U. S. NRC, Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Mr. Ronald A. Kucera, Deputy Director Department of Natural Resources P. O. Box 176 Jefferson City, Missouri 65102 Mr. Glenn L. Koester Vice President - Nuclear Kansas Gas and Electric Company 201 North Market Street Post Office Box 208 Wichita, Kansas 67201
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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-8, 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 4-1L). Since the margin to overfill for Case 1 is only 91 cubic feet, even a small increase in break flow would resul.t 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 separation. 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 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
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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 space 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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