ML090210403
| ML090210403 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/03/2009 |
| From: | Justin Poole Plant Licensing Branch III |
| To: | Meyer L Florida Power & Light Energy Point Beach |
| Poole Justin/DORL/LPL3-1/ 301-415-2048 | |
| References | |
| TAC ME0170, TAC ME0171 | |
| Download: ML090210403 (3) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 3, 2009 Mr. Larry Meyer Site Vice President FPLE Point Beach 6610 Nuclear Road Two Rivers, WI 54241-9516
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO ASTRUM IMPLEMENTATION FOR LARGE-BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS (TAC NOS.
ME0170 AND ME0171)
Dear Mr. Meyer:
By letter dated November 25,2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML083330160), FPL Energy Point Beach, LLC, submitted a request to implement the Westinghouse best-estimate loss-of-coolant accident analyses methodology known as the Automated Statistical Treatment of Uncertainty Method, as documented in Westinghouse licensing topical report WCAP-16009-P-A.
The Nuclear Regulatory Commission (NRC) staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on January 9, 2009, it was agreed that you would provide a response within 30 days of the date of this letter.
The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2048.
Sincerely, Justin C. Poole, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301
Enclosure:
Request for Additional Information cc w/encl: Distribution via ListServ
REOUEST FOR ADDITIONAL INFORMATION POINT BEACH NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301
- 1)
Compared to the current licensing basis, large-break loss-of-coolant accident (LOCA) results (obtained from a 2007 revision of the Point Beach Updated Final Safety Analysis Report), the limiting unit's best-estimate 95th percentile peak cladding temperature (PCT) is reduced by approximately 150°F. This result prompts two conclusions on the part of the staff: (1) That the power uprate case is not bounding of the current licensed thermal power case, as shown by a lower PCT, or (2) that the change in statistical sampling has redefined uncertainties in such a manner that the changes in safety margin resulting from the power uprate are masked by the method implementation. Provide information to quantify:
(A)
The PCT/MLO trend associated with the power uprate:
Compare the results of several nominal (i.e., cases whose PCTs approach the so" percentile) Point Beach WCOBRAITRAC runs assuming the implementation of an Extended Power Uprate to cases assuming the current licensed thermal power level holding other parametric assumptions the same.
(B)
The PCT/MLO trend associated with changing from COD to ASTRUM:
Compare generic results for PCT and MLO for plants that have implemented a COD/ASTRUM change without significant accompanying changes in plant parameters. Include central indicators and bounding values.
- 2)
Page 2 of Enclosure 1 to the license amendment request mentions "rackup items."
Please define this term, list the items, and provide a disposition relative to the ASTRUM analysis.
- 3)
Figures 13 and 15 bear different titles but appear to show the same information. Please explain.
- 4)
Figure 13 presents the average channel collapsed liquid level from the limiting PCT case.
Four hundred twenty to 500 seconds following the break, the oscillations in collapsed liquid level become quite severe. Please evaluate the result and explain the phenomenon. Also provide graphs from other cases in the 420-500 second time range to show whether these oscillations are anomalous for the limiting case.
- 5)
Provide reference linear heat rates assumed in the analyses: average, peak, hot rod average, hot assembly average, and hot assembly peak.
- 6)
The assumed amount of steam generator tube plugging has changed in an apparently non-conservative direction. Please explain.
Enclosure
February 3, 2009 Mr. Larry Meyer Site Vice President FPLE Point Beach 6610 Nuclear Road Two Rivers, WI 54241-9516
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO ASTRUM IMPLEMENTATION FOR LARGE-BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS (TAC NOS.
ME0170 AND ME0171)
Dear Mr. Meyer:
By letter dated November 25,2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML083330160), FPL Energy Point Beach, LLC, submitted a request to implement the Westinghouse best-estimate loss-of-coolant accident analyses methodology known as the Automated Statistical Treatment of Uncertainty Method, as documented in Westinghouse licensing topical report WCAP-16009-P-A.
The Nuclear Regulatory Commission (NRC) staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on January 9, 2009, it was agreed that you would provide a response within 30 days of the date of this letter.
The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2048.
Sincerely, IRAJT. Wengert for Justin C. Poole, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301
Enclosure:
Request for Additional Information cc w/encl: Distribution via ListServ DISTRIBUTION:
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