ML15112A601
| ML15112A601 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/13/1998 |
| From: | Labarge D NRC (Affiliation Not Assigned) |
| To: | Tuckman M DUKE POWER CO. |
| References | |
| TAC-M99349, TAC-M99350, TAC-M99351, NUDOCS 9804160065 | |
| Download: ML15112A601 (6) | |
Text
Mr. M. S. Tuckman Arl1,19 Executive Vice President Nuclear Generation Duke Energy Corporation P. 0. Box 1006 Charlotte, NC 28201-1006
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - REVIEW OF OCONEE NUCLEAR STATION UNITS 1, 2, AND 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) CHAPTER 15 TRANSIENT ANALYSIS METHODOLOGY, DPC-NE-3005-P (TAC NOS. M99349, M99350, AND M99351)
Dear Mr. Tuckman:
By letter dated July 30, 1997, Duke Energy Corporation (DEC) submitted Topical Report DPC-NE-3005-P, "UFSAR Chapter 15 Transient Analysis Methodology," for NRC review and approval. It describes the DEC methodology for analyzing the non-Loss-of-Coolant Accident UFSAR Chapter 15 transients and accidents for the Oconee Nuclear Station.
In order to continue its review of the document, the staff has determined that additional information is needed as described in the enclosure.
Sincerely, ORIGINAL SIGNED BY:
David E. LaBarge, Senior Project Manager Project Directorate 11-2 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287
Enclosure:
As stated cc w/encl: See next page Distribution:
Docket File JZwolinski LKopp LPlisco, RII PUBLIC HBerkow CLiang COgle, Rll PD 11-2 Rdg.
LBerry OGC E. Weiss DLaBarge ACRS DOCUMENT NAME: G:\\OCONEE\\OCO99349.RAI To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE PDk%
PDll-2/1-A J3W PDi -2I NAME DL48 e:cn Lerry r/Hgerg*0 DATE
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 13, 1998 Mr. M. S. Tuckman Executive Vice President Nuclear Generation Duke Energy Corporation P. 0. Box 1006 Charlotte, NC 28201-1006
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - REVIEW OF OCONEE NUCLEAR STATION UNITS -1, 2, AND 3 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR) CHAPTER 15 TRANSIENT ANALYSIS METHODOLOGY, DPC-NE-3005-P (TAC NOS. M99349, M99350, AND M99351)
Dear Mr. Tuckman:
By letter dated July 30, 1997, Duke Energy Corporation (DEC) submitted Topical Report DPC-NE-3005-P, "UFSAR Chapter 15 Transient Analysis Methodology," for NRC review and approval. It describes the DEC methodology for analyzing the non-Loss-of-Coolant Accident UFSAR Chapter 15 transients and accidents for the Oconee Nuclear Station.
In order to continue its review of the document, the staff has determined that additional information is needed as described in the enclosure.
Sincerely, David E. LaBarge, Senior Project Manager Project Directorate 11-2 Division of Reactor Projects - I/Il Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287
Enclosure:
As stated cc w/encl: See next page
Oconee, Nuclear Station cc:
Mr. Paul R. Newton Mr. J. E. Burchfield Legal Department (PBO5E)
Compliance Manager Duke Energy Corporation Duke Energy Corporation 422 South Church Street Oconee Nuclear Site Charlotte, North Carolina 28242 P. 0. Box 1439 Seneca, South Carolina 29679 J. Michael McGarry, Ill, Esquire Winston and Strawn Ms. Karen E. Long 1400 L Street, NW. -
.Assistant Attorney General Washington, DC 20005 North Carolina Department of Justice Mr. Robert B. Borsum P.0. Box 629 Framatome Technologies Raleigh, North Carolina 27602 Suite 525 1700 Rockville Pike L. A. Keller Rockville, Maryland 20852-1631 Manager - Nuclear Regulatory Licensing Manager, LIS Duke Energy Corporation NUS Corporation 526 South Church Street 2650 McCormick Drive, 3rd Floor Charlotte, North Carolina 28242-0001 Clearwater, Florida 34619-1035 Mr. Richard M. Fry, Director Senior Resident Inspector Division of Radiation Protection U. S. Nuclear Regulatory North Carolina Department of Commission Environment, Health, and 7812B Rochester Highway Natural Resources Seneca, South Carolina 29672 3825 Barrett Drive Raleigh, North Carolina 27609-7721 Regional Administrator, Region II U. S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth Street, S.W., Suite 23T85 Atlanta, Georgia 30303 Max Batavia, Chief Bureau of Radiological Health South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201 County Supervisor of Oconee County WaShalla, South Carolina 29621
REQUEST FOR ADDITIONAL INFORMATION TOPICAL REPORT DPC-NE-3005-P Please explain why there are two options available for input of the nodal axial power profile to the computer code VIPRE-01.
a) What differences result between the use of a linear interpolation compared to a spline fit?
b) How does the user determine which option to use?
- 2. -
When will the power hot channel factor -and the local-heat flux hot channel factor be input to VIPRE-01 to calculate the departure from nucleate boiling ratio (DNBR) in a subchannel? How will this cause the results of a DNBR calculation to differ from what is currently obtained?
- 3.
The Babcock and Wilcox (B&W) Correlation (BWC) and the BWC mixing vane critical heat flux correlations have been previously approved for use in VIPRE-01. Why are these considered additional features to be added to VIPRE-01/MOD2F?
- 4.
Briefly describe the enhanced iteration logic used by VIPRE-01 to converge to a minimum DNBR (MDNBR) limit.
- 5.
The moderator (boron) dilution accident for Oconee is analyzed only from conditions of power operation (Mode 1) and refueling (Mode 6). Provide analyses for Modes 2 through 5 and demonstrate that the results conform to those specified in Standard Review Plan 15.4.6.
- 6.
The acceptance criterion for the Oconee analysis of the rod ejection accident is that the offsite dose will be less than 100 percent of the 10 CFR Part 100 limits. However, NRC Regulatory Guide 1.77 and Standard Review Plan 15.4.8 specify that calculated doses should be well within 10 CFR Part 100 limits, where "well within" is defined as 25 percent of the 10 CFR Part 100 exposure guideline values. Please modify the Oconee dose acceptance criterion accordingly.
- 7.
Section 1.3 - Describe the Emergency Operating Procedures (EOPs) available at Oconee that will manually start the Emergency Feedwater (EFW) System to backup the nonsafety grade equipment in mitigating the loss of reactor coolant flow transient.
Confirm that the operator actions could be taken in time to bound the results of the analysis.
- 8.
Section 1.3 - It is stated that for certain failures in the safety grade EFW System, credit is taken for realigning EFW flow through the nonsafety Main Feedwater System and this design aspect has been reviewed and approved by the NRC. Provide the documentation for this issue.
Enclosure
-2
- 9.
Section 1.3 - One of the turbine trip circuitry channels has a slower response time than the value assumed in the analysis methodology. Confirm that the required modification will be completed prior to the approval of the proposed methodology.
- 10.
Section 1.3 - Confirm that the use of the nonsafety grade turbine trip circuitry in the transient analysis is consistent with the Oconee current licensing basis.
- 11.
Section 1.3 - It is stated that the capability to remotely throttle certain nonsafety grade valves (including the steam generator drain lines) is credited in the analysis
-methodology. Identify the areas that are affected by this assumption and-confirm that they are within the Oconee current licensing basis.
- 12.
Section 2.2 - It is indicated that the advanced solution scheme and correlations of the RETRAN-3D computer code are used in the proposed analysis methodology. Provide further discussions on how the proposed methodology could be approved without a detailed review of the RETRAN-3D code.
- 13.
Section 9.3 -Assuming a single failure of the pump monitor trip, will Cases 3 and 5 become more limiting than Cases 2 and 4 due to a reactor trip from flux/flow?
- 14.
Section 9.3.1.4 - Discuss the assumed delay time of the reactor trip.
- 15.
Section 10 (Locked Rotor) - Provide a revised analysis methodology to incorporate the following:
a) Assume a loss of offsite power with this event.
b) Include peak system pressure as a part of the acceptance criteria for this event.
c)
Since a flux/flow is the trip function for this event, a single failure of the pump monitor trip becomes a nonlimiting failure. Identify the most limiting single failure for this event.
- 16.
Section 12.0 (Turbine Trip) - It is indicated that no credit is taken for EFW flow in this event since the peak reactor coolant system (RCS) pressure will occur prior to the EFW actuation. The staff does not agree with this approach. We will require an analysis that shows there will not be a second peak pressure higher than the first peak during this transient. With an insufficient EFW flow rate, the RCS pressure could become the problem later in the transient. Also, the concern of the solid pressurizer should be addressed during the longer-term with respect to EFW flow.
- 17.
Section 13.0 (Steam Generator Tube Rupture (SGTR)) - Provide the results of a revised analysis methodology to incorporate the following:
-3 a) Assume a loss of offsite power with this event.
b) Assume a stuck-open atmospheric dump valve to maximize the radiological consequences.
- 18.
Section 13.0 - Discuss the consequences of the SGTR event assuming the nonsafety grade pressurizer heater and spray systems become inoperable.
- 19.
Section 13.0 - Discuss the EOPs available at Oconee that affect the following:
a) - Operator isolation of the ruptured steam generator following the SGTR event.
b) Prevention of steam generator overfill, assuming the maximum EFW flow rate.
- 20.
Section 15.0 (Large Steamline Break (SLB)) - Provide discussion in the following areas:
a) Why is the SLB with loss of offsite power very similar to a loss of RCS flow event.
Should an SLB with rapid RCS cooldown lead to a more severe DNBR transient?
b) Should a low initial pressurizer level lead to a lower transient pressure and lower DNBR?
c)
If a single failure of the EFW control valve is assumed, will a second MDNBR occur later into the transient due to further cooldown of the RCS?
- 21.
Section 16.0 (Small SLB) - Provide discussion in the following areas:
a) Explain why the acceptance criteria allow fuel failure and the offsite dose within 100 percent of 10 CFR Part 100 limits for a small SLB (including an inadvertent opening of a main steam relief valve), which is an incident of moderate frequency.
b) Discuss the consequences of the event assuming failure of the nonsafety grade main feedwater system and EFW is needed.