ML092880237

Issuance of Amendment No. 224, Revise Technical Specification 2.1.1.1, Departure from Nucleate Boiling Ratio (DNBR) Safety Limit

ML092880237
Person / Time
Site:	Waterford
Issue date:	11/03/2009
From:	Kalyanam N Plant Licensing Branch IV
To:	Entergy Operations
Kalyanam N, NRR/DORL/LP4, 415-1480
References
TAC ME1424
Download: ML092880237 (14)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 November 3, 2009 Vice President, Operations Entergy Operations, Inc.

Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - ISSUANCE OF AMENDMENT RE: REVISION TO THE DEPARTURE FROM NUCLEATE BOILING RATIO SAFETY LIMIT (TAC NO. ME1424)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 224 to Facility Operating License No. NPF-38 for the Waterford Steam Electric Station, Unit 3. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated June 3, 2009, as supplemented by letters dated September 22 and October 6,2009.

The amendment modifies the departure from nucleate boiling ratio (DNBR) safety limit in TS 2.1.1.1, "DNBR," based upon the Combustion Engineering 16x16 Next Generation Fuel design and the associated departure from nucleate boiling correlations.

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

Sincerely, N. Kalyanam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-382

Enclosures:

1. Amendment No. 224 to NPF-38

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY OPERATIONS, INC.

DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 224 License No. NPF-38

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

A. The application for amendment by Entergy Operations, Inc. (EOI),dated June 3, 2009, as supplemented by letters dated September 22 and October 6, 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.

Enclosure 1

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.2 of Facility Operating License No. NPF-38 is hereby amended to read as follows:

2. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 224, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented after the current cycle (Cycle 16) is completed and prior to the operation of Cycle 17.

FOR THE NUCLEAR REGULATORY COMMISSION Michael 1. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License No. NPF-38 and Technical Specifications Date of Issuance: November 3, 2009

ATTACHMENT TO LICENSE AMENDMENT NO. 224 TO FACILITY OPERATING LICENSE NO. NPF-38 DOCKET NO. 50-382 Replace the following pages of the Facility Operating License and Appendix A Technical Specifications with the attached revised pages. The revised PClges are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Operating License REMOVE INSERT

-4 Technical Specifications REMOVE INSERT 2-1 2-1

- 4 or indirectly any control over (i) the facility, (ii) power or energy produced by the facility, or (iii) the licensees of the facility.

Further, any rights acquired under this authorization may be exercised only in compliance with and subject to the requirements and restrictions of this operating license, the Atomic Energy Act of 1954, as amended, and the NRC's regulations. For purposes of this condition, the limitations of 10 CFR 50.81, as now in effect and as they may be subsequently amended, are fully applicable to the equity investors and any successors in interest to the equity investors, as long as the license for the facility remains in effect.

(b) Entergy Louisiana, LLC (or its designee) to notify the NRC in writing prior to any change in (i) the terms or conditions of any lease agreements executed as part of the above authorized financial transactions, (ii) any facility operating agreement involving a licensee that is in effect now or will be in effect in the future, or (iii) the existing property insurance coverages for the facility, that would materially alter the representations and conditions, set forth in the staff's Safety Evaluation enclosed to the NRC letter dated September 18, 1989. In addition, Entergy Louisiana, LLC or its designee is required to notify the NRC of any action by equity investors or successors in interest to Entergy Louisiana, LLC that may have an effect on the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I 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 or incorporated below:

1. Maximum Power Level EOI is authorized to operate the facility at reactor core power levels not in excess of 3716 megawatts thermal (100% power) in accordance with the conditions specified herein.

2. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 224, and the Environmental Protection Plan contained in Appendix S, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

AMENDMENT NO. 224

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE 2.1.1.1 The DI\JBR of the reactor core shall be maintained greater than or equal to 1.24.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the DNBR of the reactor has decreased to less than 1.24, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

PEAK FUEL CENTERLINE TEMPERATURE 2.1.1.2 The peak fuel centerline temperature shall be maintained less than 5080°F (decreasing by 58°F per 10,000 MWD/MTU for burnup and adjusting for burnable poisons per CENPD-382 P-A.)

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the peak fuel centerline temperature has equaled or exceeded 5080°F (decreasing by 58°F per 10,000 MWD/MTU for burnup and adjusting for burnable poisons per CENPD-382 P-A), be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4, and 5 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

WATERFORD - UNIT 3 2-1 AMENDMENT NO. 12,181,188,224

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 224 TO FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS, INC.

WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382

1.0 INTRODUCTION

By letter dated June 3, 2009 (Reference 1), as supplemented by letters dated September 22 and October 6, 2009 (References 2 and 13), Entergy Operations, Inc. (Entergy, the licensee),

proposed changes to the Technical Specifications (TSs) for Waterford Steam Electric Station, Unit 3 (Waterford 3). The proposed TS changes would reduce the departure from nucleate boiling ratio (DNBR) safety limit in TS 2.1.1.1, "DNBR," based on the Combustion Engineering (CE) 16x16 Next Generation Fuel (NGF) design and the associated departure from nucleate boiling (DNB) correlations. Specifically, the safety limit of DNBR (SLDNBR) would be reduced from ~ 1.26 with the CE-1 critical heat flux (CHF) correlation to ~ 1.24 with the WSSV-T (side supported mixing vane) and ABB-NV (non-mixing vane) CHF correlations.

The supplemental letters dated September 22 and October 6, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, did not change the U.S. Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on July 14, 2009 (74 FR 34047).

The SLDNBR in the current TS 2.1.1.1 is required to be greater than or equal to 1.26. This limit was based on the CE-1 correlation that was developed for the CHF test data applicable to CE's 16x16 fuel design using the TORC code and statistical combination of uncertainties methods.

In 2007, Westinghouse proposed and NRC approved the I'JGF assembly design (Reference 6).

The NRC-approved NGF design includes improving fuel reliability to resolve grid-to-rod fretting failures, improving fuel performance for high duty operation, and providing enhanced thermal margin for DNB. The NGF design improves heat transfer performance of the fuel design through the following design changes: (1) the addition of intermediate flow mixer (IFM) grids in the fuel assembly; and (2) the addition of side-supported mixing vanes on both the mid grids and IFM grids. For the NGF design, the WSSV-T correlation is used for DNBR calculations in the mixing vane region of the core, while the ABB-NV correlation is used to calculate the DNBR values in the hot channels in the non-mixing vane region of the core.

Enclosure 2

-2 Waterford 3's current fuel cycle (Cycle 16) has a transitional core, which consists of a partial core of NGF assemblies, and the remaining portion of the standard CE fuel design. During the transition to NGF assemblies, Waterford 3 has not taken full credit of the enhanced operating margin that is present in the NGF design. Waterford 3 will have a complete core of NGF assemblies following the refueling outage scheduled for the fall of 2009, and would take credit of the DNB benefit of the NGF mixing vanes for operations of Cycle 17 and beyond. Therefore, Waterford 3 proposed changes to TS 2.1.1.1 with incorporation of the SLDNBR associated with the ABB-NV and WSSV-T correlations for NGF assemblies. The SLDNBR value of 1.24 is the more limiting value that was determined using either the WSSV-T or the ABBV-NV correlation.

These new correlations and the associated SLDNBR will be used in the safety analyses for the next fuel cycle (Cycle 17) of operation at Waterford 3, a plant with core protection calculators (CPCs), which uses the reactor protection system (RPS) that includes the CPCs to avoid violation of the SLDBNR during normal operation and anticipated operational occurrences (ADOs). Because of existing hardware limitations in the CPC system, the licensee proposed to retain the CE-1 correlation in the CPC system. To be consistent with the safety limit of DNBR for the CE-1 correlation, the licensee proposed to use the existing value of 1.26 for the CPC low DNBR trip setpoint and Allowable Value specified in Functional Unit 10 of TS Table 2.2-1 for NGF assemblies.

2.0 REGULATORY EVALUATION

General Design Criterion (GDC) 10, "Reactor design," of Appendix A to Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50) requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any conditions of normal operation, including the effects of ADOs. In the application of pressurized-water reactors (PWRs), the SLDNBR is established to assure compliance with SAFDLs. The SLDNBR is the DNBR, which corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur.

In 10 CFR 50.36, "Technical specifications," the NRC established its regulatory requirements related to the content of TSs. In accordance with the 10 CFR 50.36 requirements, TSs are required to include items in the following five specific categories related to station operation:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. The regulations in 10 CFR 50.36(c)(1 )(i)(A) define safety limits as "limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity." In addition, 10 CFR 50.36(c)(1 )(ii)(A) also states that "[Ilimiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded."

Following these two paragraphs of 10 CFR 50.36, the SLDNBR and CPC low DNBR setpoint for reactor trip are set in TS 2.1.1.1 and in TS Table 2.2-1, respectively, for Waterford 3. In this review, the NRC staff evaluated the effect of the change of the SLDI\IBR on safe operation to assure that Waterford 3 would remain in compliance with the requirements of GDC 10 and 10 CFR 50.36.

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3.0 TECHNICAL EVALUATION

The adequate SLDNBR and setpoint of the CPC low DNBR reactor trip are important to assure that fuel rod failure due to low DNBR would not occur during normal operation and AOOs in meeting the GDC 10 requirements. TS 2.1.1.1 specifies the required SLDNBR and TS Table 2.2-1 specifies the value of the low DNBR setpoint in the CPC for reactor trip. The licensee proposed TS 2.1.1.1 with an SLDNBR applicable to NGF assemblies in the Waterford 3 core. It also proposed to retain the existing CPC low DNBR reactor trip setpoint and Allowable Value in TS Table 2.2-1 for NGF assemblies.

3.1 Safety Limit of DNBR for the NGF Design The licensee has proposed to use the ABB-NV and WSSV-T CHF correlations to perform DNBR calculations in the safety and setpoint analyses for NGF assemblies in the core. The derivations of CHF correlations were documented in Reference 5 for ABB-NV and Reference 7 for WSSV-T. Both correlations were previously approved by the NRC for use in Westinghouse's TORC and CETOP-D thermal hydraulic codes. The WSSV-T correlation is applicable to the mixing vane regions of the NGF assembly design and the ABB-NV correlation is applicable to both the standard CE fuel and the non-mixing vane regions of the NGF design. The DNBR correlation limits are 1.12 for WSSV-T and 1.13 for ABB-NV.

There are two types of approaches for calculating the minimum DNBR during normal operation conditions and AOOs: (1) the deterministic method, that assumes all adverse system parameters (such as the reactor geometry, pin-by-pin radial power distributions, inlet and exit flow boundary conditions, etc.) to occur simultaneously in the limiting subchannel; and (2) the statistical method, which involves a statistical combination of system parameter uncertainties with the CHF correlation uncertainties to determine an SLDNBR. The licensee adopted the statistical method for DNBR calculations. In support of the proposed safety limit of DNBR in TS 2.1.1.1 for the NGF design, the licensee combined uncertainties in the CHF correlation and system parameter uncertainties to determine an SLDNBR in accordance with the methods previously approved by the NRC staff and documented in Reference 4. Based on the statistical methods, the licensee determined an SLDNBR of 1.24 that assures a 95 percent probability at a 95 percent confidence level that DNB would not occur during normal operation and AOOs. This statistical DNBR limit protects the respective CHF correlation safety limits that are 1.12 for the WSSV-T correlation and 1.13 for the ABB-NV correlation.

The derivation of the SLDNBR of 1.24 relies on the methods documented in the following NRC-approved topical reports (TRs): (1) CENPD-161-P-A, TORC Code - A Computer Code for Determining the Thermal Margin of a Reactor Core; (2) CEN-356(V)-P-A, Revision 01-P-A, Modified Statistical Combination of Uncertainties; (3) CENPD-387-P-A, Revision 000, ABB Critical Heat Flux Correlations for PWR Fuel; (4) WCAP-16500-P-A, Revision 0, CE 16x16 Next Generation Fuel Core Reference Report; and (5) WCAP-16523-P-A, Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes.

During the course of the review, the NRC staff requested the licensee to address compliance with the conditions or limitations listed in the NRC safety evaluation reports (SERs) approving the TRs that allow application of the computer codes, statistical combination of uncertainties,

-4 and the ABB-NV and WSSV-T CHF correlations to Waterford 3 with the core containing NGF assemblies. In response (Reference 2), the licensee declared meeting all of the SER conditions and limitations within each of the SERs approving the TRs discussed above. The NRC staff concluded that all the applicable positions were previously reviewed and approved by the NRC.

The bases of the NRC acceptance of the TRs for Waterford 3 licensing applications were discussed in: Reference 8 for CENPD-161-P-A; Reference 9 for CEN-356(V)-P-A; and Reference 10 for CENPD-387-P-A, WCAP-16500-P-A, and WCAP-16523-P-A. Therefore, the NRC staff concludes that the use of the above noted TRs remained acceptable.

Since the proposed SLDNBR of ~1.24 was derived using the NRC-approved methods, the NRC staff concludes that it is acceptable for use of the WSSV-T and ABB-NV correlations to perform DNBR calculations in the safety and setpoint analyses for mixing vane and non-mixing vane regions of the NGF assemblies in the core, respectively.

3.2 Limiting Safety System Setting Reactor Trip Setpoint DNBR - Low The CPC system initiates the low DNBR and high local power density trips of the RPS to assure that the DNBR of the most limiting fuel assembly in the reactor core is not less than the SLDNBR specified in TS 2.1.1.1 and the fuel centerline temperature of the most limiting fuel assembly in the core does not exceed the limits specified in TS 2.1.1.2.

Because of existing CPC hardware limitations, the licensee proposed that it would retain in the CPC algorithm the CE-1 correlation, which is an NRC-approved CHF correlation with the associated SLDNBR of 1.26 for the CE standard fuel (Reference 11). Accordingly, the licensee proposed to retain the DNBR-Low trip setpoint and Allowable Value of 1.26, listed in FUNCTIONAL UNIT 10 of TS Table 2.2-1. The CPC power adjustment addressable constant, BERR1, is calculated using the WSSV-T and ABB-NV correlations in accordance with the methodology described in an NRC-approved TR (Reference 4). The BERR1 constant is calculated such that a CPC trip at a DNBR of 1.26 using the CE-1 CHF correlation assures that the bounding SLDNBR of 1.24 for the WSSV-T and ABB-NV correlations will not be exceeded during normal operations and ADOs to at least a 95 percent probability with a 95 percent confidence level. This digital setpoints process was previously approved by the NRC (Reference 6) with applicable Condition 5 on the use of the process as follows.

5. To compensate for the NRC staff concerns related to the digital setpoints process, an interim margin penalty of 6 percent must be applied to the final addressable constants (e.g., BERR1*1.06, [(1+EPOL2)*1.06 -1.0])

calculated following the 1/64 hypercube setpoints process.... Removal of this interim margin penalty will be considered after the digital setpoints methods have been formalized, documented ... , and approved by the NRC ....

In a response addressing compliance with Condition 5, the licensee stated that it will apply the 6 percent interim margin penalty to the resultant addressable constants until its removal has been approved by the NRC.

The final digital setpoints method, which will address the removal of the interim margin penalty, is documented in TR WCAP-16500-P, Supplement 1, Revision 1 (Reference 12) that is currently

-5 under the NRC staff review. In support of its application of the TR, the licensee made the following commitment:

WCAP-16500 Supplement 1 Revision 1 Safety Evaluation Report limitations or conditions will be evaluated and how they are met will be documented in the implementation package of the revision to the COLSS [Core Operating Limit Support System] and CPC setpoints and the cycle specific COLR [Core Operating Limits Report].

As stated above, the licensee's scheduled completion date for the above commitment is after the WCAP-16500, Supplement 1, Revision 1, SER is issued and prior to use as part of the reload process.

The NRC staff concluded that the licensee's response to the above cited Condition 5 and the associated commitment adequately addressed the condition imposed by the NRC on the use of the current digital setpoints method for determining the setpoint of the CPC low DNBR trip signal. Therefore, the NRC staff concludes that the proposed use of the digital setpoint method and the CPC low DNBR trip setpoint is acceptable.

3.3 Summary Based on its review, the NRC staff concludes that (1) the proposed SLDNBR for NGF assemblies and (2) the CPC low DNBR setpoint for reactor trip (listed in TS Table 2.2-1) were calculated in accordance with the NRC-approved methodologies. Therefore, the NRC staff concludes that there is reasonable assurance that the Waterford 3 has an adequate SLDNBR and CPC low DNBR setpoint for reactor trip to protect fuel rods from failure. In addition, the NRC staff concludes that Waterford 3 will continue to meet the GDC 10 requirement regarding SAFDLs and satisfy the 10 CFR 50.36 requirements regarding safety limits and limiting safety system settings in assuring safe operation of nuclear power plants. Since the proposed TS 2.1.1.1 presented in Reference 1 and FUNCTIONAL UNIT 10 in existing TS Table 2.2-1 for Waterford 3 adequately reflected the acceptable SLDNBR, and CPC low DNBR reactor trip setpoint and associated Allowable Value for the NGF assemblies, respectively, the NRC staff concludes that the proposed TS 2.1.1.1 and FUNCTIONAL UNIT 10 in existing TS Table 2.2-1 are acceptable. In addition, the staff concludes that the licensee has provided adequate justification to support the requested changes and reasonable assurance that Waterford 3 will be able to comply with the regulatory requirements and, therefore, meets 10 CFR 50.36. Therefore, the !\IRC staff concludes that the proposed TS changes are acceptable.

4.0 REGULATORY COMMITMENT In its supplemental letter dated September 22, 2009, the licensee made the following regulatory commitment:

WCAP-16500 Supplement 1 Revision 1 Safety Evaluation Report limitations or conditions will be evaluated and how they are met will be documented in the implementation package of the revision to the COLSS and CPC setpoints and the cycle specific COLR.

-6 The licensee has scheduled to complete this commitment when the SER for WCAP-16500-P-A, Supplement 1, Revision 1 is issued and prior to use as part of the reload process. The NRC staff concludes that the commitment is acceptable.

5.0 STATE CONSULTATION

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

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on July 14, 2009 (74 FR 34047). Accordingly, the amendment meets 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 the amendment.

7.0 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

8.0 REFERENCES

1. J. A. Kowalewski, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "License Amendment Request to Revise the Departure from Nucleate Boiling Ratio (Df\lBR) Safety Limit, Waterford Steam Electric Station, Unit 3, Docket No.

50-382, License No. I\IPF-38," dated June 3, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML091560027).

2. K. J. Christian, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information for the License Amendment Request to Revise the Departure from Nucleate Boiling Ratio (DNBR) Safety Limit, Waterford Steam Electric Station (Waterford 3), Docket No. 50-382, License No. NPF-38," dated September 22,2009 (ADAMS Accession No. ML092680063).

3 CENPD-161-P-A, "TORC Code - A Computer Code for Determining the Thermal Margin of a Reactor Core," dated April 1986.

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4. CEN-356(V)-P-A, Revision 01-P-A, "Modified Statistical Combination of Uncertainties,"

dated May 1988.

5. CENPD-387-P-A, Revision 000, "ABB Critical Heat Flux Correlations for PWR Fuel,"

dated May 2000.

6. WCAP-16500-P-A, Revision 0, "CE 16 x 16 Next Generation Fuel Core Reference Report," dated August 2007.

7. WCAP-16523-P-A, "Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes," dated August 2007.

8. Letter from N. Kalyanam, U.S. Nuclear Regulatory Commission, to J. Venable, Entergy Operations, Inc., "Waterford Steam Electric Station - Unit 3, Issuance of Amendment Re:

Extended Power Uprate (TAC No. MC1355)," dated April 15, 2005 (ADAMS Accession No. ML051030068).

9. Letter from C. Patel, U.S. Nuclear Regulatory Commission, to R. Barkhurst, Entergy Operations, Inc., "Issuance of Amendment No. 102 to Facility Operating License NPF Waterford Steam Electric Station, Unit 3 (TAC No. M90204)," dated March 1, 1995 (ADAMS Accession No. ML021780077).

10. Letter from N. Kalyanam, U.S. Nuclear Regulatory Commission, to Entergy Operations, Inc., "Waterford Steam Electric Station, Unit 3 - Issuance of Amendment RE: Request to Support Next Generation Fuel; Review and Approval of Revised Emergency Core Cooling System (ECCS) Performance Analysis; and Review and Approval of Supplement to the ECCS Performance Analysis (TAC Nos. MD6954, MD6363, and MD6954)," dated April 15,2008 (ADAMS Accession No. ML080880014).

11. CENPD-162-P-A, "C-E Critical Heat Flux," dated September 1976; Supplement 1-A, dated February 1977; and CENPD-207-P-A, "C-E Critical Heat Flux Part 2 Non-uniform Axial Power Distribution," dated December 1984.

12. WCAP-16500-P, Supplement 1, Revision 1, "Application of CE Methodology for CE 16X16 Next Generation Fuel (NGF)," dated October 2008 (ADAMS Accession No. ML083050498, Proprietary Information. Not publicly available).

13. K. Christian, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Revised Page for License Amendment Request to Revise the Departure from Nucleate Boiling Ratio (DNBR) Safety Limit, Waterford Steam Electric Station (Waterford 3),

Docket No. 50-382, License No. NPF-38," dated October 6,2009 (ADAMS Accession No. ML092800514).

Principal Contributor: S. Sun Date: November 3, 2009

November 3, 2009 Vice President, Operations Entergy Operations, Inc.

Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - ISSUANCE OF AMENDMENT RE: REVISION TO THE DEPARTURE FROM NUCLEATE BOILING RATIO SAFETY LIMIT (TAC NO. ME1424)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 224 to Facility Operating License No. NPF-38 for the Waterford Steam Electric Station, Unit 3. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated June 3, 2009, as supplemented by letters dated September 22 and October 6,2009.

The amendment modifies the departure from nucleate boiling ratio (DNBR) safety limit in TS 2.1.1.1, "DNBR," based upon the Combustion Engineering 16x16 Next Generation Fuel design and the associated departure from nucleate boiling correlations.

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

Sincerely, IRA!

N. Kalyanam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-382

Enclosures:

1. Amendment No. 224 to NPF-38

2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC RidsNrrDorlLpl4 Resource RidsOgcRp Resource LPLIV r/f RidsNrrDorlDpr Resource RidsRgn4MailCenter Resource RidsAcrsAcnw_MailCTR Resource RidsNrrPMWaterford Resource SSun, NRR/DSS/SRXB RidsNrrDirsltsb Resource RidsNrrLA.IBurkhardt Resource AAttard, NRR/DSS/SNPB ADAMS Accession No ML092880237 OFFICE NRRlLPL4/PM NRRlLPL4/LA DSS/SRXB/BC DSS/SNPB/BC OGC NRRlLPL4/BC NRR/LPL4/PM MMarkley NAME NKalyanam JBurkhardt GCranston AMendiola BHarris CFLyon for NKalyanam DATE 10/16/09 10/16/09 10/7/09 10/30/09 10/26/09 11/3/09 11/3/09 OFFICIAL RECORD COpy