ML032240121
| ML032240121 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/19/2003 |
| From: | Padovan L NRC/NRR/DLPM/LPD3 |
| To: | Denise Wilson Nuclear Management Co |
| References | |
| TAC MB7185 | |
| Download: ML032240121 (6) | |
Text
August 19, 2003 Mr. David L. Wilson Site Vice President Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT REQUEST FOR ADDITIONAL INFORMATION RELATED TO REVISED LONG-TERM CONTAINMENT RESPONSE AND NET-POSITIVE SUCTION HEAD ANALYSES (TAC NO. MB7185)
Dear Mr. Wilson:
The Nuclear Management Company, LLCs (NMCs), December 6, 2002, application requested that the U.S. Nuclear Regulatory Commission (NRC) approve proposed changes to the Updated Safety Analysis Report for the Monticello Nuclear Generating Plant. The NRC staff is reviewing your request and finds that additional information is needed as shown in the enclosed Request for Additional Information (RAI).
I discussed the enclosed RAI with Mr. R. Loeffler of your organization on August 7, 2003. We agreed that NMC will respond to the RAI within 30 days of receipt of this letter. Please contact me at (301) 415-1423 if you have questions or need to revise this date.
Sincerely,
/RA/
L. Mark Padovan, Project Manager, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-263
Enclosure:
Request for Additional Information cc w/encl: See next page
ML032240121 OFFICE PDIII-1/PM PDIII-1/LA PDIII-1/SC NAME MPadovan RBouling LRaghavan DATE 08/15/03 08/14/03 08/19/03
Monticello Nuclear Generating Plant cc:
Jonathan Rogoff, Esquire General Counsel Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 U.S. Nuclear Regulatory Commission Resident Inspectors Office 2807 W. County Road 75 Monticello, MN 55362 Manager, Regulatory Affairs Monticello Nuclear Generating Plant Nuclear Management Company, LLC 2807 West County Road 75 Monticello, MN 55362-9637 Robert Nelson, President Minnesota Environmental Control Citizens Association (MECCA) 1051 South McKnight Road St. Paul, MN 55119 Commissioner Minnesota Pollution Control Agency 520 Lafayette Road St. Paul, MN 55155-4194 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Commissioner Minnesota Department of Health 717 Delaware Street, S. E.
Minneapolis, MN 55440 Douglas M. Gruber, Auditor/Treasurer Wright County Government Center 10 NW Second Street Buffalo, MN 55313 Commissioner Minnesota Department of Commerce 121 Seventh Place East Suite 200 St. Paul, MN 55101-2145 Adonis A. Neblett Assistant Attorney General Office of the Attorney General 445 Minnesota Street Suite 900 St. Paul, MN 55101-2127 John Paul Cowan Executive Vice President & Chief Nuclear Officer Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Nuclear Asset Manager Xcel Energy, Inc.
414 Nicollet Mall, R.S. 8 Minneapolis, MN 55401 August 2003
ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO REVISED LONG-TERM CONTAINMENT RESPONSE AND NET-POSITIVE SUCTION HEAD ANALYSES NUCLEAR MANAGEMENT COMPANY, LLC (NMC)
MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 The Nuclear Regulatory Commission (NRC) staff requests the following additional information related to NMCs December 6, 2002, application:
- 1. (a) What assurance is there that the K value will remain at 147 or above?
(b) How often does NMC verify this?
(c) Has NMC made a measurement to verify that the K value is currently greater than 147?
- 2. If NMC has revised the calculation of residual heat removal room temperature from the analysis provided in NMCs March 4, 1997, letter to the NRC, briefly describe the changes and the conclusions.
- 3. Briefly describe the analysis that concludes that the piping temperature limit can be increased to 196.7 degrees F.
- 4. Describe the SAFER/GESTR models and the assumptions used to calculate Monticellos response to a vessel isolation with high-pressure coolant injection (HPCI) unavailable.
Include a nodalization diagram. Describe any conservatism in this analysis.
- 5. Regarding Section 4.5 of General Electrics report GE-NE-0000-0002-8817-01, R1, dated September 2002, Monticello Nuclear Generating Plant Long-term Containment Analysis, explain how it is physically possible to have a service water temperature of 94 degrees F and a suppression pool temperature of 90 degrees F under steady state conditions.
- 6. Section 4.4 of GE-NE-0000-0002-8817-01, R1, page 4-12, begins by discussing the design-basis loss-of-coolant accident (LOCA) analysis with the updated heat exchanger K value and updated data. What are these updated data?
- 2 -
- 7. Verify that the information in the table below is correct.
Break Size Residual Heat Removal (RHR)
Heat Exchanger K
Service Water Temp
F Peak Suppression Pool (SP)
Temp Comment Large break*
143.1**
90 195.6 Direct SP cooling Large break 147 90 194.1 Direct SP cooling Large break 147 90 194.2 Containment spray cooling Large break 147 94***
195.8 Direct SP cooling Large break 147 94 196.5 Containment spray cooling Reactor isolation 143.1 90 194.0 One RHR loop, HPCI unavailable, direct SP cooling Reactor isolation 143.1 90 167.0 Two RHR loops with HPCI unavailable, direct SP cooling
.01 ft2 143.1 90 190.0 One RHR loop with HPCI unavailable, direct SP cooling
.1 ft2 143.1 90 191.2 One RHR loop with HPCI unavailable, direct SP cooling
- A study of single failures in the June 19, 1997, NMC application showed the failure of one emergency diesel generator with loss of offsite power to be most limiting.
- As stated above, the original K value for the RHR heat exchanger is 143.1 BTU/sec-F while the updated value is 147 BTU/sec-F.
- SP water temperature remains at 90 degrees F.
- 7. Describe how heat transfer to structures is modeled for the net positive suction head calculations.
- 8. In Exhibit F, Figures 8, 9, 10, and 11, NMC showed required and available overpressure for the isolation event and the Appendix R event. What is the source of the pressure for these events since the steam from the safety/relief valves is condensed in the suppression pool?
- 3 -
- 10. Verify that the table below is correct.
Accident Scenario Current Licensing Basis Value Proposed Change to Licensing Basis Value Peak containment pressure
((short-term large-break (LB)
LOCA))
Date: 7/26/96 Power: 1880 megawatts thermal (Mwt) 40 psig Unchanged Peak containment temperature (short-term LB LOCA)
Date: 7/26/96 Power: 1880 Mwt 331 degrees F Unchanged Peak bulk pool temp (long-term LB LOCA)
Date: 6/19/97 Power: 1880 Mwt 194.2 degrees F Date: 12/6/02 Power: 1775 Mwt 195.6 degrees F Max local pool temperature (short-term LB LOCA)
Date: 7/26/96 Power: 1880 Mwt 194 degrees F Unchanged Drywell wall temperature (small steam line break)
Date: 7/26/96 Power: 1880 Mwt 273 degrees F Unchanged Reactor isolation peak pool temperature None Date: 12/6/02 Power: 1775 Mwt 194 degrees F
- 11. Verify that there has been no change in Monticellos licensing basis for calculating the debris loading on the emergency core cooling system suction strainers.
- 12. What value of required net position suction head used for the calculation of required containment overpressure?
- 13. Regarding Exhibit F, describe, or reference, how the effects of pipe friction are accounted for, including the increase to account for aging?