NRC 2009-0042, Supplement to License Amendment Request 258, Incorporate Best Estimate Large Break Loss of Coolant Accident (LBLOCA) Analyses Using Astrum

Supplement to License Amendment Request 258, Incorporate Best Estimate Large Break Loss of Coolant Accident (LBLOCA) Analyses Using Astrum

ML091000170
Person / Time
Site:	Point Beach
Issue date:	04/08/2009
From:	Meyer L Florida Power & Light Energy Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2009-0042
Download: ML091000170 (5)

Text

FPL Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241 FPL Energy.

Point Beach Nuclear Plant April 8, 2009 NRC 2009-0042 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Docket 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Supplement to License Amendment Reauest 258 Incorporate Best Estimate Large Break Loss of Coolant Accident (LBLOCA)

Analvses Using ASTRUM

References:

(1) FPL Energy Point Beach LLC, Letter to NRC dated November 25,2008, License Amendment Request 258, Incorporate Best Estimate Large Break Loss of Coolant Accident (LOCA) Analyses Using ASTRUM (ML083330160)

(2) NRC Electronic Mail to FPL Energy Point Beach LLC, dated March 25, 2009, Draft RAI questions from Technical Specifications Branch on LBLOCAITSTF 363 Amendment (ML090840481)

FPL Energy Point Beach, LLC submitted Point Beach Nuclear Plant (PBNP) Units Iand 2, proposed License Amendment Request 258 for Commission review and approval pursuant to 10 CFR 50.90 (Reference 1). The amendment requests use of NRC-approved WCAP-I 6009-P-A, Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM), and revises Technical Specification (TS) 5.6.4.b to reference WCAP-16009-P-A. The request also proposes to implement Technical Specification Task Force (TSTF) Traveler-363A in order to eliminate the revision numbers and dates from the list of topical reports identified in TS 5.6.4.b.

Reference 2 identified that a statement required to fully implement TSTF-363A was omitted in the marked up pages for TS 5.6.4.b that was provided to the NRC as part of the original submittal (Reference 1).

The enclosure to this letter submits the revised TS 5.6.4.b marked up pages which incorporate the omitted statement.

An FPL Group company

Document Control Desk Page 2 This supplement does not alter the proposed changes, the significant hazards consideration or the environmental considerations previously provided in Reference (1).

In accordance with 10 CFR 50.91, a copy of this application, with enclosure, is being provided to the designated State of Wisconsin Official.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on April 8, 2009.

Very truly yours, FPL Energy P n c h , LLC site-vice-president Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

ENCLOSURE FPL ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST 258 INCORPORATE BEST ESTIMATE LARGE BREAK LOSS OF COOLANT ACCIDENT (LOCA) ANALYSES USING ASTRUM MARKED-UP TECHNICAL SPECIFICATION PAGES 2 pages follow

5.6.4 CORE OPERATING LIMITS REPORT (COLR) (continued)

(8) L C 0 3.2.3, "Axial Flux Difference (AFD)"

(9) L C 0 3.3.1, "Reactor Protection System (RPS) lnstrumentation -

Overtemperature AT"

( I 0) L C 0 3.3.1, "Reactor Protection System (RPS) Instrumentation -

Overpower AT"

( I 1) L C 0 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits" (12) L C 0 3.9.1, "Boron Concentration" The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. When an initial assumed power level of 102 percent of the original rated thermal power is specified in a previously approved method, 100.6 percent of uprated rated thermal power may be used only when the main feedwater flow measurement (used as the input for reactor thermal output) is provided by the Caldon leading edge flowmeter (LEFM) as described in reports IIand 12 listed below. When main feedwater flow measurements from the LEFM are unavailable, a power measurement uncertainty consistent with the instruments used shall be applied.

Future revisions of approved analytical methods listed in this Technical Specification that currently reference the original Appendix K uncertainty of 102 percent of the original rated thermal power should include the condition given above allowing use of 100.6 percent of uprated rated thermal power in the safety analysis methodology when the LEFM is used for main feedwater flow measurement.

The COLR will contain the comnlete identification for each of the TS referenced topical renorts used to nrepare the COLR (i.e., report number, title. revision. date. and anv sunplements).

The approved analytical methods are described in the following documents:

1) WCAP-I 4449-P-A, "Application of Best Estimate Large Break LOCA Methodology to Westinghouse

.. PWR's with Upper Plenum injection^^ ? , (cores containing I 422V+ fuel)

(2) WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology" ddyW4% I (3) WCAP-I 1397-P-A, "Revised Thermal Design Procedure; Apt4 I Point Beach 5.6-3 Unit 1 - Amendment No.

Unit 2 - Amendment No.

5.6.4 CORE OPERATING LIMITS REPORT (COLR) (continued)

(4) WCAP-I 4787-P, Re+-& "Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wisconsin Electric Power Company Point Beach Units I& 2 (Fuel Upgrade &

Uprate to 1656 MWt-NSSS Power with Feedwater Venturis, or 1679 MWt-NSSS Power with LEFM on Feedwater Header)",

(5) m O 5 4 - P - A , "Westinghouse Small Break ECCS Evaluation Model Using The NOTRUMP Code,"

(6) WCAP-I 0054-P-A, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code:

Safety Injection into the Broken

. . :Loop and COSl Condensation Model;'--A&h&w 2, ! ? e ~ w r ? , JtAy4QW (7) WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions; (8) WCAP-10216-P-A, . . "Relaxation of Constant Axial Offset Control," aewster,??,, -F (9) WCAP-10924-P-A, "Large Break LOCA Best Estimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection; ,-

-(cores not containing 422 V+ fuel)

( I 0) WCAP-I 0924-P-A, "LBLOCA Best Estimate Methodology:

Model Description and Validation: Model Revisions; Ve4ww4, A . (cores not containing 422 V+ fuel)

(11) =c.:' E-eport-80 "TOPICAL REPORT:

Improving Thermal Power Accuracy and Plant Safety While Increasing Operating

.. Power Level Using the LEFMJTM System?. -

(12) Caldon, Inc., Engineering Report-IGOP, "Supplement to Topical Report ER-80P: . . Basis for a Power Uprate With the LEFMJTM System; (13) WCAP-16009-P-A, "Realistic Larae-Break LOCA Evaluation Methodoloav Usina the Automated Statistical Treatment of Uncertaintv Method (ASTRUM)"

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Point Beach 5.6-4 Unit I- Amendment No. 3a.r Unit 2 - Amendment No. 242