ML15112A426
| ML15112A426 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/17/1999 |
| From: | Labarge D NRC (Affiliation Not Assigned) |
| To: | Tuckman M DUKE POWER CO. |
| References | |
| TAC-MA4713, TAC-MA4714, TAC-MA4715, NUDOCS 9903240050 | |
| Download: ML15112A426 (5) | |
Text
March 17, 1999 Mr. M. S. Tuckman Distribution:
CHawes COgle, Rll Executive Vice President C~okeit Eilidl DLaBarge Nuclear Generation PUBLIC RCaruso Duke Energy Corporation PD 11-2 RF WJensen P. 0. Box 1006 (ECO7H)
JZwolinski/SBlack OGC Charlotte, NC 28201-1006 HBerkow ACRS
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - TRANSIENT ANALYSIS METHODOLOGY, TOPICAL REPORT DPC-NE-3005P - OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 (TAC NOS. MA4713, MA4714, AND MA4715)
Dear Mr. Tuckman:
By letter dated February 1, 1999, Duke Energy Corporation (Duke) submitted Revision 1 of Topical Report DPC-NE-3005P, that describes the new Duke methodology for analyzing the Oconee Updated Final Safety Analysis Report Chapter 15 Non-Loss-of-Coolant Accident transients and accidents.
The principal issue remaining is that credit for main feedwater system isolation would no longer be assumed in the analysis of main steam line breaks. Isolation of main feedwater is not accomplished with safety related equipment. The staff, therefore, required that analyses be performed assuming continued feedwater runout. The licensee performed the requested analyses using the RETRAN code that was previously approved by the staff for Oconee.
In order to complete its review, the staff needs additional information, as shown in the enclosure. We request that you respond by April 16, 1999, as discussed with Mr. Gregg Swindlehurst of your staff.
Sincerely, Original signed by" David E. LaBarge, Senior Project Manager Project Directorate 11-2 Division Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287
Enclosure:
Request for Additional Information cc w/encls: See next page Document Name: G:\\OCONEE\\04713RAI.WPD To receive a copy of this document, indicate in the box:
"C" = Copy without attachmentlenclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE PDIkPKJ, I PDII-2/LA PDl H
NAME D gecn CHawesIT1h HB rka-'
DATE 3
7 /99 3 / l/99
/ 7/99 OFFICIAL RECORD COPY 9903240050 990317 PDR ADOCK 05000269
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 17, 1999 Mr. M. S. Tuckman Executive Vice President Nuclear Generation Duke Energy Corporation P. 0. Box 1006 (ECO7H)
Charlotte, NC 28201-1006
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - TRANSIENT ANALYSIS METHODOLOGY, TOPICAL REPORT DPC-NE-3005P - OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 (TAC NOS. MA4713, MA4714, AND MA4715)
Dear Mr. Tuckman:
By letter dated February 1, 1999, Duke Energy Corporation (Duke) submitted Revision 1 of Topical Report DPC-NE-3005P, that describes the new Duke methodology for analyzing the Oconee Updated Final Safety Analysis Report Chapter 15 Non-Loss-of-Coolant Accident transients and accidents.
The principal issue remaining is that credit for main feedwater system isolation would no longer be assumed in the analysis of main steam line breaks. Isolation of main feedwater is not accomplished with safety related equipment. The staff, therefore, required that analyses be performed assuming continued feedwater runout. The licensee performed the requested analyses using the RETRAN code that was previously approved by the staff for Oconee.
In order to complete its review, the staff needs additional information, as shown in the enclosure. We request that you respond by April 16, 1999, as discussed with Mr. Gregg Swindlehurst of your staff.
Sincerely, David E. LaBarge, Senior Project Manager Project Directorate 11-2 Division Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287
Enclosure:
Request for Additional Information cc w/encls: See next page
Oconee Nuclear Station Mr. J. E. Burchfield cc:
Compliance Manager Ms. Lisa F. Vaughn Duke Energy Corporation Legal Department (PBO5E)
Oconee Nuclear Site Duke Energy Corporation P. 0. Box 1439 422 South Church Street Seneca, South Carolina 29679 Charlotte, North Carolina 28201-1006 Ms. Karen E. Long Anne Cottington, Esquire Assistant Attorney General Winston and Strawn North Carolina Department of 1400 L Street, NW.
Justice Washington, DC 20005 P. 0. Box 629 Raleigh, North Carolina 27602 Mr. Rick N. Edwards Framatome Technologies L. A. Keller Suite 525 Manager - Nuclear Regulatory 1700 Rockville Pike Licensing Rockville, Maryland 20852-1631 Duke Energy Corporation 526 South Church Street Manager, LIS Charlotte, North Carolina 28201-1006 NUS Corporation 2650 McCormick Drive, 3rd Floor Mr. Richard M. Fry, Director Clearwater, Florida 34619-1035 Division of Radiation Protection North Carolina Department of Senior Resident Inspector Environment, Health, and U. S. Nuclear Regulatory Natural Resources Commission 3825 Barrett Drive 7812B Rochester Highway Raleigh, North Carolina 27609-7721 Seneca, South Carolina 29672 Mr. Steven P. Shaver Virgil R. Autry, Director Senior Sales Engineer Division of Radioactive Waste Management Westinghouse Electric Company Bureau of Land and Waste Management 5929 Carnegie Blvd.
Department of Health and Environmental Suite 500 Control Charlotte, North Carolina 28209 2600 Bull Street Columbia, South Carolina 29201-1708 Mr. William R. McCollum, Jr.
Vice President, Oconee Site County Supervisor of Oconee County Duke Energy Corporation Walhalla, South Carolina 29621 P. 0. Box 1439 Seneca, SC 29679
Request for Additional Information Topical Report DPC-NE3005-P, Revision 1 UFSAR Chapter 15 Transient and Accident Methodology Oconee Nuclear Station, Units 1, 2, and 3 The following questions deal with the revised RETRAN analyses of a postulated main steam line break (MSLB) described in Section 15 of DPC-NE-3005-P. The analyses include the consequences of failures to isolate main feedwater.
- 1. Pages 15-3 and 15-4 state that an inlet core mixing fraction is of 46 percent was used with the MSLB analyses and that the mixing fraction was obtained from tests performed on Oconee Unit 1. Discuss the mixing tests and how the mixing fraction was derived. Justify that this fraction is appropriate for MSLB analysis.
- 2. Page 15-4 states that the RETRAN transport delay model was disabled for the MSLB analyses since flow in the intact loop becomes stagnant so that use of the model is inappropriate. The staff understands that enthalpy transport was also disabled in the affected coolant loop that would not become stagnant. Consideration of enthalpy transport in the affected coolant loop might lead to an increased rate of core overcooling. Provide the results of an evaluation showing the effect of considering enthalpy transport in the affected loop on core power and the departure from nucleate boiling ratio (DNBR).
- 3. A postulated MSLB with off site power available would permit continued main feedwater flow and reactor coolant pump operation. Page 15-10 states that the reactor coolant pumps in the unaffected loop are assumed to trip at 100 seconds. This was stated to be conservative since reverse heat transfer was occurring in the steam generator, which was minimizing overcooling. Operators are trained to trip all reactor coolant pumps on loss of subcooling margin. Evaluate the MSLB for the case of continued main feedwater flow but with operator action to trip all reactor coolant pumps. Determine the effect on DNBR of any return to power.
- 4. During a large main steam line break at Oconee the core flood tanks (CFTs) would be expected to discharge and add boric acid, which would act to reduce core power production.
As the CFTs discharge, the cover gas will cool to a temperature lower than the liquid.
RETRAN does not have a non-equilibrium CFT model. To assume equilibrium might lead to a higher CFT pressure and a greater discharge than would actually occur. Justify that the CFT model in RETRAN is appropriate for analysis of main steam line breaks.
- 5. Page 15-12 states that the method of calculating boron reactivity feedback involves use of the average core boron concentration and the average core boron worth. Boric acid from the CFTs and safety injection would reach the bottom of the core first, which would have a lower reactivity importance than the average core. Justify that using the average core boron concentration and average core boron worth is appropriate.
- 6. Page 15-14 states that the limiting assumption with respect to maximizing overcooling and reactivity addition has been determined by analysis to be the case when the Integrated
-2 Control System (ICS) is controlling main feedwater. This case was found to cause more overcooling than the case that assumes failure of the ICS, which would permit unlimited main feedwater addition to the affected steam generator. Provide comparisons of the conditions within the affected steam generator secondary that would cause the ICS operating case to be the worst case. Compare steam generator level, pressure, temperature, and heat transfer coefficients that are calculated by the RETRAN code.
- 7. To enable the NRC staff to perform audit calculations and sensitivity analyses if needed, please provide electronic copies (on computer disks) of the RETRAN and VIPRE input decks used for MSLB analysis.