ML092931684

Issuance of Amendment Use of ASTRUM for LBLOCA

ML092931684
Person / Time
Site:	Point Beach
Issue date:	10/29/2009
From:	Justin Poole Plant Licensing Branch III
To:	Meyer L Point Beach
Poole Justin/DORL/LPL3-1/ 301-415-2048
References
TAC ME0170, TAC ME0171
Download: ML092931684 (16)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 29, 2009 Mr. Larry Meyer Site Vice President NextEra Energy Point Beach, LLC Point Beach Nuclear Plant 6610 Nuclear Road Two Rivers, WI 54241

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: ASTRUM IMPLEMENTATION FOR LARGE-BREAK LOCA ANALYSIS (TAC NOS. ME0170 AND ME0171)

Dear Mr. Meyer:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No.235 to Renewed Facility Operating License No. DPR-24 and Amendment No. 239 to Renewed Facility Operating License No. DPR-27 for the Point Beach Nuclear Plant, Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated November 25, 2008, as supplemented by letters dated March 4, April 8, and September 15, 2009.

These amendments implement the Westinghouse best-estimate loss-of-coolant accident analysis methodology known as the Automated Statistical Treatment of Uncertainty Method, as documented in Westinghouse licensing topical report WCAP-16009-P-A.

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

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

Enclosures:

1. Amendment No. 235 to DPR-24

2. Amendment No. 239 to DPR-27

3. Safety Evaluation cc w/encls: Distribution via ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FPL ENERGY POINT BEACH, LLC DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 235 License No. DPR-24

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by FPL Energy Point Beach, LLC (the licensee),

dated November 25, 2008, as supplemented by letters dated March 4, April 8, and September 15, 2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations setforth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

- 2

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 4.B of Renewed Facility Operating License No. DPR-24 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 235,are hereby incorporated in the renewed operating license. FPLE Point Beach shall operate the facility in accordance with Technical Specifications.

3.

This license amendment is effective as of the date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of issuance:

October 29 I 2009

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FPL ENERGY POINT BEACH, LLC DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 239 License No. DPR-27

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by FPL Energy Point Beach, LLC (the licensee),

dated November 25, 2008, as supplemented by letters dated March 4, April 8, and September 15, 2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

-2

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 4.B of Renewed Facility Operating License No. DPR-27 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 239, are hereby incorporated in the renewed operating license. FPLE Point Beach shall operate the facility in accordance with Technical Specifications.

3.

This license amendment is effective as of the date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

-X9/P~'

Robert J. Pascarelli, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of issuance: October 29, 2009

ATTACHMENT TO LICENSE AMENDMENT NO. 235 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 AND LICENSE AMENDMENT NO. 239 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 DOCKET NOS. 50-266 AND 50-301 Replace the following pages of the Facility Operating Licenses and Appendix A Technical Specifications (TS) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT License Pages License Pages DPR-24, page 3 DPR-24, page 3 DPR-27, page 3 DPR-27, page 3 TS Page TS Page 5.6-4 5.6-4

- 3 D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, FPLE Point Beach to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E.

Pursuant to the Act and 10 CFR Parts 30 and 70, FPLE Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.

4.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A.

Maximum Power Levels FPLE Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1540 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 235, are hereby incorporated in the renewed operating license.

FPLE Point Beach shall operate the facility in accordance with Technical Specifications.

C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.

Renewed License No. DPR-24 Amendment No. 235

- 3 C. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, FPLE Point Beach to receive, possess and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed source for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, FPLE Point Beach to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E.

Pursuant to the Act and 10 CFR Parts 30 and 70, FPLE Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.

4.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Levels FPLE Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1540 megawatts thermal.

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 239 are hereby incorporated in the renewed operating license.

FPLE Point Beach shall operate the facility in accordance with Technical Specifications.

C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.

Renewed License No. DPR-27 Amendment No. 239

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 CORE OPERATING LIMITS REPORT (COLR) (continued)

(4)

WCAP-14787-P, Rev. 2, "Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wisconsin Electric Power Company Point Beach Units 1 & 2 (Fuel Upgrade &

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

October, 2002 (approved by NRC Safety Evaluation, November 29,2002).

(5)

WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using The NOTRUMP Code," August 1985.

(6)

WCAP-10054-P-A, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code:

Safety Injection into the Broken Loop and COSI Condensation Model," Addendum 2, Revision 1, July 1997.

(7)

WCAP-8745-P-A, "Design Bases for the Thermal Overpower

~T and Thermal Overtemperature ~T Trip Functions,"

September 1986.

(8)

WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control," Revision 1A, February 1994.

(9)

WCAP-10924-P-A, "Large Break LOCA Best Estimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection," and Addenda, December 1988. (cores not containing 422 V+ fuel)

(10)

WCAP-10924-P-A, "LBLOCA Best Estimate Methodology:

Model Description and Validation: Model Revisions," Volume 1, Addendum 4, August 1990. (cores not containing 422 V+ fuel)

(11) Caldon, Inc., Engineering Report-80P, "TOPICAL REPORT:

Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM,rTM System," Revision 0, March 1997.

(12) Caldon, Inc., Engineering Report-160P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate With the LEFM,rTM System," Revision 0, May 2000.

(13) WCAP-16009-P-A "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM) " Revision O. January 2005.

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 1 - Amendment No.~, 235 Unit 2 - Amendment No.~, 239

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 235 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 AND AMENDMENT NO. 239 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 FPL ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301

1.0 INTRODUCTION

By application to the U.S. Nuclear Regulatory Commission (NRC, Commission) dated November 25, 2008, as supplemented by letters dated March 4, April 8, and September 15, 2009 (Agencywide Documents Access and Management System (ADAMS)

Accession Nos. ML083330160, ML090771303, ML091000170, and ML092590389, respectively), FPL Energy Point Beach, LLC (the licensee), requested changes to the Technical Specifications (TSs) for the Point Beach Nuclear Plant, Units 1 and 2. The supplements dated March 4, April 8, and September 15, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on January 13, 2009 (74 FR 1714).

The proposed changes would implement the Westinghouse best-estimate loss-of-coolant accident (LOCA) analysis methodology known as the Automated Statistical Treatment of Uncertainty Method (ASTRUM), as documented in Westinghouse licensing topical report WCAP-16009-P-A (Reference 2). Specifically, the proposed changes would incorporate ASTRUM into its licensing basis large-break (LB) LOCA analysis, and add a corresponding reference to WCAP-16009-P-A into TS 5.6.4.b, Core Operating Limits Report (COLR)

References.

The original submittal letter dated November 25, 2008, included a request to implement Technical Specification Task Force (TSTF) Traveler 363-A. Following discussions with the NRC staff, the licensee submitted the September 15, 2009, letter, to withdraw that request. The NRC staff acknowledges the withdrawal of the request to implement TSTF 363-A.

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2.0 REGULATORY EVALUATION

Title 10 of the Code ofFederal Regulations (10 CFR), Part 50, Section 46, provides the requirements for the emergency core cooling system (ECCS), requiring that the ECCS must be designed such that, when analyzed using the guidance set forth in 10 CFR 50.46, it demonstrates acceptable performance subject to the criteria contained in 10 CFR 50.46(b)(1) through (b)(5).

The NRC staff reviewed the licensee's request to implement the ASTRUM methodology to ensure the following:

(1) ASTRUM is generically NRC-approved and acceptable for implementation at Point Beach.

(2) The ASTRUM analysis demonstrates acceptable ECCS performance relative to the LOCA acceptance criteria 10 CFR 50.46(b)(1) through (b)(3).

Note that the NRC staff, per Item 2 above, reviewed the licensee's LBLOCA analysis for compliance with 10 CFR 50.46(b)(1) - (b)(3). Separate analyses and operator actions are credited to demonstrate compliance with the remaining acceptance criteria, and these items are not affected by the implementation of the ASTRUM LBLOCA analysis.

3.0 TECHNICAL EVALUATION

Westinghouse obtained generic NRC approval of its original topical report describing the best estimate (BE) LBLOCA methodology in 1996, for 3 and 4-loop pressurized-water reactors. This method is known as the Code Qualification Document (CQD) methodology (Reference 5). NRC approval of the methodology is documented in the NRC safety evaluation report appended to the topical report. This methodology was later extended to 2-loop Westinghouse plants with Upper Plenum Injection (UPI) in 1999, as documented in the NRC safety evaluation report appended to the UPI topical report (Reference 6).

Westinghouse recently completed a program to revise the statistical approach used to develop the Peak Cladding Temperature (PCT) and oxidation results at the 95th percentile. This method is based on the CQD methodology and follows the steps in the Code Scaling Applicability and Uncertainty (CSAU) methodology [NUREG/CR-5249]. However, the uncertainty analysis (Element 3 in CSAU) is replaced by a technique based on order statistics. The ASTRUM methodology replaces the response surface technique with a statistical sampling method in which the uncertainty parameters are simultaneously sampled for each case. The approved ASTRUM evaluation model is documented in WCAP-16009-P-A (Reference 2).

3.1 Summary of Licensee's Analysis and Comparison to Current Licensing Basis The licensee's current licensing basis LOCA analysis is based on the 1996 BE-LBLOCA methodology, and was analyzed for Point Beach at the current licensed thermal power level of 1540 megawatts-thermal (MWt). The licensee stated that the current licensing basis LBLOCA analysis, applicable to both units, demonstrated a PCT of 2131 of.

-3 The proposed ASTRUM analysis was performed for Point Beach operating at an assumed extended power uprate power level of 1800 MWt, and showed a limiting PCT of 1875°F for Point Beach Unit 1. The licensee stated that FPL Energy Point Beach and its vendor, Westinghouse Electric Company LLC, continue to have ongoing processes which ensure that LOCA analysis input values conservatively bound current operating values.

The full results are tabulated below for Unit 1; the Unit 2 limiting results were less severe and hence not presented in this Safety Evaluation Report.

Parameter ASTRUM Results 10 CFR 50.46 Limits Peak Cladding Temperature 1975°F 2200°F Local Metal Oxidation 2.61%

17%

Core-Wide Oxidation 0.386%

1%

3.2 NRC Staff Evaluation of Licensee's Analytic Results The NRC staff reviewed the information submitted by the licensee and concluded that the ASTRUM method is NRC-approved to analyze LBLOCAs at two-loop, upper plenum injection plants such as Point Beach, and that the licensee's analysis demonstrates acceptable performance relative to the 10 CFR 50.46 acceptance criteria at uprated conditions. In consideration of these items, the NRC staff finds the licensee's request to implement ASTRUM acceptable. The NRC staffs finding is based not only on the considerations discussed above, but also on evaluation of the licensee's response to requests for additional information (RAls),

as discussed in the following subsections of this safety evaluation report.

RAI 1: Demonstrate PCT Effects of the Methodology Improvement and Power Uprate The NRC staff requested additional information because the results presented by the licensee reflected a power uprate, and did not reflect the plant's actual operating state (Reference 1). In response, the licensee quantified the PCT effect of the power uprate by analyzing several LOCA cases in the 50th-percentile PCT range to demonstrate that the uprate causes a predicted increase in PCT on the order of 100°F (Reference 4).

The licensee also demonstrated the effect that the improved statistical modeling has on the best-estimate PCT by comparing generic PCT trends for plants replacing the COO method with ASTRUM. These results showed that the ASTRUM statistical modeling approach can improve (reduce) the PCT by approximately 500°F (Reference 4).

The PCT sensitivities demonstrate that, while the ASTRUM analysis assumes a power uprate and still indicates a lower PCT than the current, COD-based licensing basis analysis, it is still credible. This is because the statistical improvements affect a greater decrease in the PCT than the increase resulting from the assumed power uprate.

The NRC staff finds the licensee's response acceptable because it quantifies the conservatism claimed by assuming a higher power level than that at which the plant currently operates. The NRC staff agrees that assuming implementation of a power uprate provides LBLOCA results that are conservative relative to a LBLOCA model that assumes the current licensed power level.

- 4 RAI2: Define "Rackup Items" The licensee's analysis refers to "rackup items" included in the results for the current licensing basis (Reference 1), and the NRC staff requested that the licensee define this term, list the items, and clarify their disposition relative to the proposed ASTRUM analysis (Reference 3).

The licensee responded in Reference 4, stating that rackup items are plant-specific PCT assessments against the LBLOCA analysis of record, and are tracked so that the aggregate PCT is known relative to the PCT indicated by the analysis of record. The rackup items reflect changes in plant configuration and the correction of modeling errors. The licensee listed the current rackup items and clarified that, upon implementation of the new BE-LBLOCA analysis method will be eliminated. The NRC staff finds the licensees response acceptable because it clarifies information contained in the application and does not affect the acceptability of the licensee's analysis or results.

RAI 3: Provide Corrected Figures The NRC staff reviewed the figures and tables included in Reference 1 and determined that some figures bore captions and titles that were inconsistent with the information presented in the figures. The licensee addressed the NRC staffs concern by providing corrected figures in Reference 4. The NRC staff reviewed the figures, and finds the licensee's response acceptable because the corrected figures provided clarifying information that did not affect the acceptability of the licensee's analysis or results.

RAI4: Evaluate Collapsed Liquid Level Oscillations The NRC staff reviewed the licensee's LBLOCA analytic results and observed that there were sharp oscillations in the collapsed liquid level toward the end of the transient (Reference 1). The NRC staff was concerned that these oscillations could be the result of a numerical imbalance in the limiting case calculation (Reference 3).

The licensee addressed the NRC staffs concern by examining the collapsed liquid level from several other limiting cases (Reference 4). The licensee noted that the oscillations observed by the !\\IRC staff appeared in these cases also. The licensee also stated that the oscillations were the result of a hydrostatic imbalance in the reactor vessel arising from lateral flow redistribution, and provided plots of vertical continuous liquid mass flow rate compared to collapsed liquid level to support this claim. The plots demonstrate the relationship between horizontal and vertical flow in the vessel, and the collapsed liquid level.

The plots show that flow communication between the upper plenum and the core region is, in part, the cause of the flow oscillations. The vertical liquid mass, when decreasing (indicative of increasing flow in the downward direction), results in a liquid level increase. As a result, it can be seen that the flow redistributes laterally, as shown in the figures showing horizontal continuous liquid mass flow.

The results and explanation provided by the licensee indicate that the flow oscillations are calculated thermal-hydraulic phenomena and not numerical instabilities. The NRC staff accepts the licensee's response because it resolves the NRC staff's concern by indicating that the analytic results represent predictable flow phenomena.

-5 RAI 5: Provide Linear Heat Rates The NRC staff requested that the licensee provide the linear heat rates assumed in the LBLOCA analysis (Reference 3). The licensee's response provided the linear heat rates requested by the NRC staff and confirmed that the linear heat rates were analyzed in a manner consistent with the ASTRUM methodology (Reference 4). The assumed peak linear heat rate in the limiting PCT case for the hot rod is 16.1 kW1ft, which is a value based on sampling the core average linear power and the COLR maximum total peaking factor value.

The licensee's response indicated that the peak linear heat rate was determined in a manner consistent with the ASTRUM methodology document, and is based on expected average linear heat rates and limiting core peaking factors. The NRC staff finds, therefore, that the assumed linear heat rates are acceptable for the LBLOCA analysis.

RAI6: Clarify Assumptions Regarding Steam Generator Tube Plugging In its review of the license amendment request, the NRC staff observed that the current licensing basis LBLOCA analysis assumes 25-percent steam generator tUbe plugging, and the proposed ASTRUM LBLOCA analysis assumes 1O-percent steam generator tUbe plugging (Reference 1).

The NRC staff requested that the licensee address the difference in assumed steam generator tube plugging (Reference 3).

In response, the licensee explained that 25-percent was overly conservative, and that 10-percent remained an acceptable amount of analyzed steam generator tube plugging, because the maximum number of steam generator tubes plugged is in Point Beach Unit 1, Steam Generator B, with 6 plugged tubes, which amounts to 0.19-percent (Reference 4).

Therefore, the 10-percent plugging assumption remains conservative relative to plant operation.

Because the licensee clarified that the 1O-percent steam generator tube plugging assumption remains conservative relative to operation at the plant, the NRC accepts the licensee's response.

3.3 Summary The NRC staff reviewed the licensee's analytic results that support the requested implementation of ASTRUM for analyzing the LBLOCA events. Based on its review of the data that the licensee provided, the NRC staff requested additional information to clarify the analytic assumptions and results, to which the licensee responded with adequate supplemental information.

Based on two considerations: (1) the ASTRUM methodology is NRC-approved for modeling the LBLOCA at a two-loop, upper plenum injection plant; and (2) the Point Beach-specific results show acceptable ECCS performance relative to the 10 CFR 50.46 acceptance criteria, the NRC staff finds that the licensee's request to implement ASTRUM for LBLOCA analysis at Point Beach is acceptable.

-6

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Wisconsin State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

These amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or change a surveillance requirement. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously published a proposed finding that these amendments involve no significant hazards consideration and there has been no public comment on such finding (74 FR 1714). Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Ben Parks, NRR Date: October 29,2009

Mr. Larry Meyer Site Vice President NextEra Energy Point Beach, LLC Point Beach Nuclear Plant 6610 Nuclear Road Two Rivers, WI 54241 SUB~IECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: ASTRUM IMPLEMENTATION FOR LARGE-BREAK LOCA ANALYSIS (TAC NOS. ME0170AND ME0171)

Dear Mr. lVIeyer:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 235 to Renewed Facility Operating License No. DPR-24 and Amendment No. 239 to Renewed Facility Operating License No. DPR-27 for the Point Beach Nuclear Plant, Units 1 and 2, respectively.

The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated November 25, 2008, as supplemented by letters dated March 4, April 8, and September 15, 2009.

These amendments implement the Westinghouse best-estimate loss-of-coolant accident analysis methodology known as the Automated Statistical Treatment of Uncertainty Method, as documented in Westinghouse licensing topical report WCAP-16009-P-A.

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRA!

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

Enclosures:

1. Amendment No. 235 to DPR-24

2. Amendment No. 239 to DPR-27

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